ML20133G886

Tenth Partial Response to First Set of QA Interrogatories & Request to Produce.Qa Program Complies W/Criteria of 10CFR50,App B.Supporting Documents & Certificate of Svc Encl.Related Correspondence

ML20133G886
Person / Time
Site:	Braidwood
Issue date:	10/09/1985
From:	COMMONWEALTH EDISON CO.
To:	ROREM, B.
References
CON-#485-778 OL, NUDOCS 8510160172
Download: ML20133G886 (235)

Text

.. _. _ _ _ _ _ _ _ _ . _ . _ . _ _ .. __.__ _ _ . . _ . . . _ _ _

171 fa Cctcher 9, 1985 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION l p * * *gDENCE

~

l I In the Matter of: )

CC 0 C0fet0NWEALTH EDISON COMPANY ) ^'

) ) Docket Nos. 50-4560 b j (Braidwood ) 50-457 j Station, Units 1 and 2) )

• 85 (CI 19 A10 :28 I

l APPLICANT'S TENTH PARTIAL RESPONSE TO OFFICE or SEW ito-I ROREN'S FIRST SET OF QUALITY ASSURANCE ghgC h INTERROGATORIES AND REQUEST TO PRODUCE On July 2, 1985, Intervenors Rorem, et al. ("Intervenors"),

flied their First Set of Qaality Assurance Interrogatories and Request to Produce. On July 30, 1985 Cosmonwealth Edison Company (" Applicant")

flied objections to certain of those discovery requests and provided to

{ Intervonors' counsel a partial response to Intervenors' discovery i

requests. Applicant's first partial response was revised and resubmitted on August 1, 1985. Applicant flied a second partial response on August j 5, 1985, a third on August 10, 1985, a fourth on August 13, 1985, a fifth i

i on August 14, 1985, a sixth on August 28, 1985, a seventh on September 6, l

! 1985, an eighth on September 12, 1985, and a ninth on September 24 I

1985. This submission, which is Applicant's tenth partial response to Intervonors' discovery requests, provides responses to requests which

, were the subject of the Licensing Board's September 27, 1985 order

compelling discovery. 'In addition, supporting affidavits are included.

i Applicant also is submitting with this pleading a list which identifies

}

documents which are being withheld by Applicant under claims of attorney-client and attorney work product privilege. Investigation l

i continues for responses to all of the requests which Applicant has l answered and Applicant will provide at a later date updated indices of '

i l

documents made available for inspection and copying.

8510160172 85 g 56 t

PDR ADOCK PDR

! Q 0130H Q3

o Response to request for homo addresses and telephone numbers of individuals previously identified in responses to Specific Interrogatories 19, 51, 58, and 59.

In accordance with an agreement reached among Applicant's counsel, Intervonors' counsel, and counsel for the NRC Staff, lists of home addresses and telephone numbers for individuals identified '

in the response to Specific Interrogatories 19, 51, 58, and 59 are being provided only to Intervenors' counsel and counsel for the NRC Staff. Distribution has been limited in this manner to protect the privacy interests of individuals to the extent possible.

I i

i

, 0313H/ October 8, 1985

\

1 i

i Specific Interrogatory 1 Do you agree that the Braiduced Quality Assurance (QA)

Program must comply with each of the criteria of Appendix B to 10 CFR Part 50 in order to establish Applicant's en-titlement to the licenses sought in this proceeding?

Response

Unless exempted or waived pursuant to NRC regulations, the Braidwood Quality Assurance Program, which is defined as the Commonwealth Edison Company Topical Report, must be found to comply with the quality assurance criteria of Appendix B to 10 C.F.R. Part 50 prior to the issuance of the operating licenses for Braidwood Station. This response is provided by Applicant's counsel, f

f f

)

O o

Specific Interrogatory 3 Does the Braidwood Quality Assurance Program comply with each of the criteria of Appendix B to 10 CFR Part 50?

Response

The Preliminary Safety Analysis Report ("PSAR"), the Fina) Safety Analysis Poport ("FSAR"), and the Safety Evalu-ation Report ("SER") define the Braidwood Quality Assurance Program as the Commonwealth Edison Company Topical Report.

This program complies with each of the criteria of 10 CPR 50, Appendix B. Documents describing the NRC reviews of the Topical Report were provided in response to Interrogatory

12. The acceptability of the most recent revision reviewed and found to continue to satisfy the requirements of 10 CFR 50, Appendix B by the NRC is contained in NRC letter of August 29, 1985 from C. J. Paperiello to C. Reed (a ttached) .

UNITED STATES

[jptacg#% NUCLEAR REGULATORY COMMisslON

[ '.

i nEcios m 6U 0 3 E j 7ss nooseveLT nono

. 0,,

CLEN EL LYN. ILUNOls 60137 AUG 291985 Docket No. 50-10; Docket No. 50-237; Docket No. 50-249 Docket No. 50-254; Docket No. 50-265 Docket No. 50-295; Docket No. 50-304 Docket No. 50-373; Docket No. 50-374 Docket No. 50-454; Docket No. 50-455 Docket No. 50-456; Docket No. 50-457 Commonwealth Edison Company ATTN: Mr. Cordell Reed Vice President Post Office Box 767 Chicago, IL 60690 Gentlemen:

Thank you for the submittal of your Quality Assurance Topical Report CE-1-A, Revision 37. We have completed our review of the changes and find that the revised program continues to satisfy the requirements of 10 CFR 50, Appendix,B and is therefore acceptable.

We appreciate your timely submittal of the infomation required by 10 CFR 50.54(a) and 50.55(f). Please contact Mr. Frank Hawkins of my office (312-790-5555) with any questions you have regarding this matter.

Sincerely,

~

y Carl J. Pap riello,' Director Division of Reactor Safety cc: Attached List for Distribution l

. AUG 291985 Commonwealth Edison Company 2 1

D. L. Farrar, Director of Nuclear Licensing .

D. J. Scott, Plant Manager N. Kalivianakis, Plant Manager K. L. Graesser, Plant Manager Jan Norris, Project Manager, NRR '

G. J. Diederich, Plant Manager i V. I. Schlosser, Project Manager Gunner Sorensen, Site Project ,

Superintendent R. E. Querto Plant Manager M. Wallace, Project Manager D. Shamblin, Construction Superintendent J. F. Gudac, Plant Manager .

C. W. Schroeder, Licensing and Compliance Superintendent '

DMB/ Document Control Desk (RIDS)

Resident Inspector, Dresden, Quad Cities, Zion, LaSalle, Byron, Braidwood ,

Phyllis Dunton, Attorney General's Office, Environmental Control Division Mayor, City of Zion D. W. Cassel, Jr. , Esq. .

Diane Chavez, DAARE/ SAFE  !

W. Paton, ELD L. Olshan, NRR LPM H. S. Taylor, Quality Assurance  :

Division  !

J. W. McCaffrey, Chief, Public -

Utilities Division i 4

G. T. Ankrum, IE l i

I 9

t-L I

l

)

Specific Interrogatory &

What are the bases for your response to Nos. 1-3? Please identify all documents, physical evidence, testimony or oral statements by any person and legal authority on which you rely in support of your position.

Response

The bases for Applicant's responses to Specific Interrogatories 1 and 3 are provided in those answers. Specific Interrogatory 2 is not applicable.

~ . . . .

0309H/ October 8, 1985

Interrogatory 5 i

Does the workmanship in the actual, fabrication, con-struction and testing of safety-related structures, systems and components meet or exceed all applicable standards?

Response

It is the objective of Commonwealth Edison to assure

that workmanship in the design, fabrication, construction I

and testing of safety-related structures, systems and com-ponents meets or exceeds all applicable standards. This objective is addressed through the implementation of a management system that acquires, develops and applies re-

sources, as necessary to produce the level of workmanship i

expected. That management system in place at Braidwood had been described previously in response to intervenor inter-rogatories 13, 14, 15, 16, 24, 51, 56 and 61.

However, it is recognized that in any project com-Y parable in size and complexity to a nuclear power plant such as Braidwood, workmanship that does not in the first in-stance meet all applicable standards will sometimes occur.

I NRC regulations themselves account for such occurrences by providing in 10 CFR 50, Appendix B, Criterion XVI - Cor-i rective Action, that the quality assurance program required i

~

in 10 CFR 50.34 to be implemented by a nuclear production facility license applicant include measures to assure that conditions adverse to quality, such as nonconforming work-

, , - - - .. ,y -

, , - ~ - .m,- - --

--g - .m-.-e- ,,,,,,~,a- , , , - , , ,,-----en-.- - -- - , - . -,-.- , - - .-

. . ~ _ . . .- -. .-.. . . _ - . = . . . -.. .- - .-- - . . - .-

l.

i i

1 i .

manship, be promptly identified and corrected. The Commonwealth A

i Edison Company Quality Assurance Program for Nuclear Generating Stations, Topical Report CE-1-A, which applies.to Braidwood e Station includes such measures in Section 16.0 (Corrective j Action) of the Topical Report. That Topical Report has been

! made available to the intervenor in response to Interrogatory

1. 12.

i 1

i 1

i

?

1

]

i i

1 i

t i

i i

i l

SPECIFIC INTERROGATORY 6

6. Please identify all sources of standards (e.g., FSAR, ASME Code) applicable to the actual design, fabrication, construction and testing of safety-related structures, systems, and components.

RESPONSE

10 CFR Section 50.55a, " Codes and Standards, the PSAR, and the FSAR are " sources of standards." The entire FSAR contains frequent references to various standards; however, certain sections of the FSAR summarize Commonwealth Edison Company's commitments to standards. These sections include:

FSAR Table 3.8-2, List of Specifications, codes, and Standards FSAR Chapter 17.0 - Quality Assurance, which incorporates by reference Commonwealth Edison Topical Report CE-1-A FSAR Appendix A - Application of NRC Regulatory Guides FSAR Appendix B - Construction Material Standards and Quality Control Procedures Commonwealth Edison Topical Report CE-1-A, Section 2.2, which provides for compliance with certain regulatory guides and ANSI standards Copies of the above referenced sections are being provided to Intervenors' counsel.

0311H/ October 8, 1985

_ _ . - _ . . - . . _ _ - _ _ _ _ . _ _ _ _ _ _ . ..._ __ _ _ _ _. _ _ _ - m._-

il Interrogatory 7 I

If the answer to No. 5 is negative, please describe in

, detail the respects in which such workmanship does not meet

or exceed all applicable standards or is indeterminate, and 2

explain fully the factual and legal basis for your answer. .

J l Response:

As was discussed in response to Interrogatory #5, the possibility of workmanship deficiencies at Braidwood Station I is contemplated, and specific provision for corrective i

action to resolve such deficiencies has been made as part of t

the quality assurance program applicable to Braidwood 4

i 4

Station. Those respects in which such workmanship did not initially meet all applicable standards or is indeterminate, and which are within the subject matter areas of the con-4 tention, have been fully explained in response to intervenor Interrogatory #58.

i l

l l

i 1

i k

I i

1 l

1 Specific Interrogatory 8 What are the bases for your responses to Nos. 5-77 please identify all documents, physical evidence, testimony, or oral statements by any person and legal authority on which you rely in support of your position.

Response

The bases for Applicant's responses to Specific It.terrogatories 5, 6, and 7 are provided in those answers, i

W

}

1 l

0314H/ October 8, 1985

l I

' I Specific Interrogatory 9 )

Please identify each deficiency in design and construction as defined in 10 CFR Section 50.55(e) and for each indicate: the classification of its significance (i.e., classified under which subsections, 50.55(e)(1)(1-iv); the 10 CFR Part 50 Appendix A, General Design Criteria, to which each relates and the respects in which it reflects noncompliance; the report number, and date, if any; the names, titles, addressses and telephone numbers of nach person responsible for the deficiency, its discovery, its reporting, and its corrective action; a detailed description of the deficiency and its safety implications; a detailed description of its corrective action.

Supplemental Response Applicant has previously responded to Specific Interrogatory 9, limited to the scope of the contention, in Applicant's first revised and seventh partial responses. In light of the Board's September 27, 1985 Order compelling a response to this interrogatory, counsel for Applicant discussed this matter further with Intervenors' counsel. Based on these discussions, counsel for Applicant believes that a complete answer already has been provided to this interrogatory as it is limited in scope by Intervenors' motion to compel and the Board's order.

l 0312H/ October 8, 1985

Specific Interrogatory 17 Please describe in detail the circumstances and procedures, if any, under which Quality Control inspection criteria may be waived.

Response

Q.C. inspection criteria are included in specifications and drawings issued by the Architect Engineer for Braidwood Station to the contractor responsible for a particular scope of work.

The contractor is responsible for incorporating the inspection criteria into quality procedures by direct reiteration or by reference, as appropriate. The procedures are reviewed and approved by CECO Project Construction, Site Quality Assurance, and CECO Engineering personnel prior to use. The Architect Engineer also participates in the procedure reviews and is responsible for confirming that the procedure implementation will result in completed work which meets the specification or drawing inspection criteria. Formal notification of procedure approval is then forwarded by CECO to the applicable contractor.

There are no predetermined circumstances under which established Q.C. inspection criteria are waived. Currently, in circumstances where a waiver is considered, an appropriate engineering review is performed. If the review finds the waiver acceptable, the inspection criteria is waived. Under the CECO Quality Assurance Program the Field Change Request (FCR) system, the Engineering Change Notice (ECN) system, or ,

1 the Non-Conformance Report (NCR) system may be used to disposition proposed Q.C. inspection criteria waivers. One

circumstance where Q.C. inspection criteria would be considered for waiver is where the item to be inspected is inaccessible for inspection because of other completed construction.

A specific circumstance relative to the waiver of Q.C.

inspection criteria which was not dispositioned through engineering occurred in the 1982 time frame for equipment internal cleanliness and is addressed in the response to i

Interrogatory 58, Contention Item 12.G. CECO is not aware of any other situations where Q.C. inspection criteria have been i

waived without an engineering review.

i I

4 e

t t

I i i r l

I a

i l  ;

i

_ . - . --_ .,. .-_ . . _ . , ,_.,.,--,m,,.__,-,_--. -- _ _,,.., -_,___- - - _ - . . , . - . . _ _ , , - . - _ . . , , - - . . _ . . . - - , , - , , . . _ . ,

Specific Interrogatory 50 In a February 1,1984 Chicago Tribune article, NRC Region III Administrator James C. Keppler said, in part, with regard to commonwealth Edison and Braidwood: "One has to question whether the workload has become unmanageable for the staff they have, and I've raised that for management to consider. But I have to be concerned that they are spread thin at the top..." Did Mr. Keppler or the NRC raise this matter with Commonwealth Edison? Or did Edison otherwise identify such deficiencies? If so, please describe in detail the circumstances, Commonwealth Edison's response and any corrective action taken and the results. Please identify any documents which reflect the answers to these questions.

Response

Mr. James Keppler did express to Commonwealth Edison the opinion that the general office management may be spread thin. It is Edison's understanding that the comment was directed towards general office oversight of operating stations activities. Edison does not believe this has been or is currently a problem. However, over the past few years a number of changes have been made to strengthen the general office oversight of station operations. In July, 1984, the number of execut)ves reporting to the Vice President of Nuclear Operations was increased fac1 two to three with the creation of the position of Assistant Vice President for Nuclear Services. This helped to reduce the load on the General Manager of Nuclear Stations and allowed closer attention to nuclear services provided to the stations.

! Additionally in September, 1985, two Division Vice President positions were created reporting to the General Manager of Nuclear Stations. Each Division Vice President will have responsibilities for i

three of Edison's nuclear stations. This will allow greater oversight of station operations. Both Division Vice Presidents are former station managers and holders of senior reactor operator licenses. These changes 0302H/ October 8, 1985

l 1

resulted from a long term plan developed with the assistance of a management consultant which is directed towards making an efficient transition from an organization managing three operating nuclear stations to one managing six operating nuclear stations.

CA i

l i

0302H/ October 8, 1985 l

i l

Interrogatory 52 Have any quality assurance weaknesses or deficiencies -

at Braidwood been caused by management action or inaction?

If so, please describe in detail. Have any adverse per-sonnel actions (for example, te rmination, demotion, transfer or suspension) been taken by Commonwealth Edison and/or its contractors against any person (s) because of QA deficiencies or weaknesses? If so, please identify the circumstances and

persons involved, including names and addresses, and identify any documents reflecting such instances and their reso-lution.

r i

1

Response

The Commonwealth Edison management of the Braidwood Project is ultimately responsible for every action or in-action taken by individuals who are working on the Project.

While every Quality Assurance Program deficiency can be j

attributed to action or inaction on the part of an indi-1 vidual or individuals, the nature and types of deficiencies

, identified at Braidwood are not unexpected for a large nuclear construction project and are not considered in-dicative of broad based weaknesses in the Quality Assurance

Program. Overall, Commonwealth Edison management has been l aggressive in taking substantive actions in carrying out its

, responsibility to implement an effective Quality Assurance Program, and seizing opportunities for improvement.

Large nuclear construction projects, extend over a i

4 period of several years, and experience a " Project dynamics",

as certain activities ar'e more or less pervasive and signi-

)i i

. - , , . ,c. < - - . - . , . - - - - - , - . . - - - . - . . - - , . - , . . . , . . , . . - , , , , - , - , - - - , . - - . -

ND

't. s x\

NN kw ficant at any time in the Project's life. As such, it is h not uncommon for a number of personnel and organizational changes to take place during the life of the Project. Such has been the case on the Braidwood Project, particularly in the latter stages of completion when the coordination and interfacing of a number of work groups and the reviewing of a large volume of documentation for final acceptance become significant activities. A number of personnel changes have been made on the Braidwood Project, as reflected in the response to Interrogatory 51. To the best of my knowledge, the,se were not considered adverse personnel actions in terms of each individual's career advancement.

Similarly, within the contractor organizations, " Pro-ject dynamics" were also at work which led to a comple-n.entary set of personnel and organizational changes. In large measure, the personnel changes were not considered

" adverse" as defined above. However, in certain instances they were. In August, 1983, L. K. Comstock replaced an individual serving in the role of Quality Control Manager due to his ineffectiveness in meeting organizationas com-mitments. To the best of my knowledge, this was the only adverse personnel action taken, relative to Quality As-surance Program implementation, from among the more senior site contractor management personnel. As described in the Attachment to this response, the response to Interrogatory

1 l

l l

l 19 provides additional instances of termination of Quality Assurance and Quality Control personnel for performance related cases.

Commonwealth Edison management has always had, as a b primary objective at Braidwood, the effective implementation w

94 of a Quality Assurance Program in accordance with all ap-

'i' plicable regulations and requirements. To that end, manage-Sg ment actions have been taken at various times throughout the r.

'hlife of the Project to either address concerns raised within A

thq Company or by the NRC, or, to strengthen the overall effe0D4veness of the Quality Assurance Program at Braidwood.

NN' Examplek(of actions taken by Commonwealth Edison management

'x'N to improve'fhe A effectiveness of the Quality Assurance Pro-

\'

gram can be fdgnd in the responses to Interrogatories 24 and 4

51, and more rec'9ptly \

in early 1985, with respect to the dismissal by L. K.'Nomstock of a lead Quality Control Super-visor, Mr. R. Saklak,qs a result of an investigation arising out of the charges of inb{midation and harassment of Quality Control Inspectors employed', y L. K. Comstock, s

\ 's s\

N;s

'\

w k 's

\'.

'\;

N.

'\,

'N

\

\

\

\

y .

\

T

_. . _ _ . _ _ . . ~ - .- - . . _ _ . ..- .- - - . -. -.

ATTACHMENT Intervenor's First Set, Interrogatory 19 requested the reason for termination for each person no longer employed in Quality Assurance and Quality Control for Commonwealth Edison and each contractor.

) Commonwealth Edison provided the response to this interrogatory as i follows:

Partial Response 5 -- Provided the response for Commonwealth Edison Quality Assurance personnel and for Comstock Quality Assurance and Quality Control personnel.

Partial Response 7 -- Provided the response for Quality 3

Assurance and Quality Control personnel employed by Pullman Sheet Metal and Pittsburgh Testing Laboratories, a

Partial Response 9 -- Provided the response for Quality Assurance and Quality Control personnel employed by Phillips, Getschow Company, G. K. Newberg Company, a revised list for Pittsburgh Testing Laboratories and additional information for one employee of Pullman Sheet Metal, i

4 T

l i

l l

l (1822d) l l

Interrogatory 57 In what respects are NRC requirements understood to be either minimum or maximum requirements with regard to the design and construction of Braidwood? Please explain in detail and identify any documents which reflect this answer.

Response

The NRC requirements applicable to the design and con-struction of Braidwood are delineated in 10 CFR 50, Domestic Licensing of Production and Utilization Facilities. With the exception of Appendix A to 10 CFR 50, General Criteria for Nuclear Power Plants, which indicates that the require-ments defined therein are minimum requirements, Commonwealth Edison is aware of no regulatory guidance which describes NRC requirements as either minimum or maximum requirements.

It is Commonwealth Edison's objective to meet or exceed all applicable NRC requirements including those related to design and construction except in those instances for which exemption has been justified and authorized in accordance with the provisions for specific exemption contained in 10 CFR 50.12.

With respect to 10 CFR 50, Appendix A to which re-ference was previously made, the design of Braidwood is intended to provide margins suf ficient to meet or exceed the generalized minimum requirements specified by the NRC.

However, the design process neither is nor can be defined in such detail that the extent to which NRC requirements are l

[

1 exceeded ca.r. De identified in advance. Therefore, there may be instances for which the design philosophy and, on a more microscopic basis, design margins are identical to that required by the NRC.

This is also true to construction activities at Braidwood.

In this regard, intervenors have identified certain specific NRC requirements applicable to Braidwood and to particular circumstances pertaining to the design or construction of Braidwood for which the NRC Staf f identified apparent de-ficiencies in meeting these requirements. These deficiencies relate to the quality assurance program or implementation thereof by Commonwealth Edison or certain of its contractors with respect to the design or construction of Braidwood Station. Specifically, these NRC requirements, to which Intervenors refer in the contention, are defined in 10 CFR 50, Appendix B - Quality Assurance Criteria for Nuclear I

Power Plants and Fuel Reprocessing Plants. The requirements delineated in this appendix that are enumerated by Inter-venors and for which circumstances are described in the contention are: Criterion I - Organization; - Design Con-trol; Criterion V - Instruction, Procedures and Drawings; Criterion VI - Document Control; Criterion VIII - Identi-fication and Control of Materials, Parts and Components; Criterion IX - Control of Special Processes; Criterion X -

Inspection; Criterion XV - Nonconforming Materials, Parts or i

f

_ - , _ _ .- - , -___,,.-r- - - -- r--- - -

Components; Criterion XVI - Corrective Action; Criterion XVII - Quality Assurance Records; and Criterion XVIII -

Audits.

Because of the very general wording of the criteria of Appendix B, that wording is amenable to more than one in-terpretation. With respect to the circumstances enumerated by intervenors in the contention, which are alleged to demonstrate conduct contrary to these requirements, Ap-plicant's responses to Interrogatory 58, discu'ss each of these circumstances in detail including a discussion of Applicant's interpretation of the NRC requirement in ques-tion to the extent that interpretation differed from the interpretation stated by the NRC staff. Those specific circumstances for which such difference were identified are discussed in the following listed parts of the response to Interrogatory 58: Parts 1.A., 3.A.1, 3.A.2, 3.A.3, 3.A.4.A.,

3.A.4.C, 3.C., 4.B., 6.A., 6.B.4., 6.C., 6.F., 8.D., 8.E.,

8.F., 9.A., 9.B., 9.C., 9.E., 10.B., 10.C., 10.E., 10.D.,

ll.A., ll.B., ll.C., 12.A., 12.B.3., 12.E., 12.F., 12.G.,

13.B., 14.A., and 14.B.4. For many of these cases Commonwealth Edison acceded to the NRC staff interpretation. In all instances, actions either were or will be implemented to resolve the specific concerns identified by the NRC staff.

n

.4 .,

RE1ATED CORRESPONDENCE AFFIDAVIT OF CORDELL REED COLKETED

'Mu:

I, Cordell Reed, being first duly sworn, depose and state as follows:

1) I am employed by Commonwealth Edison Company as Vic'85PrGHde%t A10 :29 in charge of Nuclear Operations.

r, r r -  ::t.i

2) My business address is 72 W. Adams St., P.O. Box 76h ichicagoi:, ,

Illinois 60690. FRANCH

3) I am responsible for the preparation of the answer to Specific Interrogatory 50 submitted in Applicant's Tenth Partial Response to Intervenor's First Set of Quality Assurance Interrogatories and Request to Produce on October 9, 1985.

4) To the best of my knowledge and belief, the statements contained in the answer to Specific Interrogatory 50 of Applicant's Tenth Partial Response to Intervenor's First Set of Quality Assurance Interrogatories and Request to Produce are true and correct.

Further affiant says not.

t

( C(G NW

-i % e-se 0 ,} \

Cordell Reed Signed nd sworn to before me

)

'this9 day of October 1985

! ' ,- 1 f Y(1 And110$

Notary Public sp it t

' u./ Ludr.i.si n Ex;!ics April 6,10f3 0306H

. . . , ..a L d

1 HELATED CORRESPONDENCf.

AFFIDAVIT OF Michael J. Wallace DOLKETED UNC I, Michael J. Wallace, being first duly sworn, depose and state g follows: 0 g

1) I an employed by Commonwealth Edison Company as Project y(q , ,- g t m Manager at Braidwood Nuclear Power Station. 00C4ETi% & SF A ' '

ERANCH

2) My business address is Braidwood Nuclear Power Station, Braceville, Illinois 60407.

3) I am responsible for the preparation of the answers to Specific Interrogatories 24 and 51 filed by Intervonors Rorem, et.al. These answers were prepared and submitted in Applicant's Third Partial Response to Intervonor's First Set of Quality Assurance Interrogatories and Request to Produce on or about August 9, 1985. More recently, I am responsible for the preparation of the answer to specific Interrogatory 52 submitted in Applicant's Tenth Partial Response on 0:tober 9, 1985.

4) To the best of my knowledge and belief, the statements contained in the answers to Specific Interrogatories 24, 51 and 52 of Applicant's Response to Intervenor's First Set of Quality Assurance Interrogatories and Request to Produce are true and correct.

Further affiant says not.

h .; , [

Af} j Y d.-

MichaelJ.pilace Signe gnd sworn to before me this 3 A day of October 1985 N -[ Wi NotaryV>ubi;,c ' /

l .

0303H

REi.ATED CORRESPONDENCE Affidavit of Charles M. Allen 00U.ETED W*

I, Charles M. Allen, being first duly sworn, depose and state as follows: '85 0CT 15 N0:29

1) I am employed by Comonwealth Edison Company as a OfF O  :.c. 4 .

• t h' U '

Licensing Engineer at Braidwood Nuclear Station. DCC'ti fj'

2) My business address is Braidwood Nuclear Power Station, Braceville, Illinois 60407.

3) I was responsible for supervising the preparation of the lists of home addresses and telephone numbers provided in Applicant's Tenth Partial Response to Rorem's First Set of Quality Assurance Interrogatories. To the best of my knowledge and belief,

' these lists are true, correct, and complete with regard to individuals identified in Applicant's response to Specific Interrogatories 19, 51, 58, and 59.

Further affiant says not.

i /

}[

Charles M. Allen i

d sworn before me Sign this g3 1 day of October 1985 l bA .

Notary / Public V My Commissien Expires !I!3 h e

1 0240H j

s RELATED CGRRESPONDENCR 00LXOEI AFFIDAVIT OF EUGENE E. FITZPATRICK 'MNRC I Eugene E. Fitzpatrick, being first duly sworn, depose andgtaM 15 A10 :29 as follows:

1) I am employed by Conunonwealth Edison Company as Assistant: 3 M !fEPW . .

t 00C%EI i Manager of Quality Assurance.

2) My business address is Braidwood Nuclear Pcwor Station, Braceville, Illinois 60407,

3) I am responsible for the preparation of the answers to Specific Interrogatories 3, 4 to the extent it relates to 3, and 17 submitted in Applicant's Tenth Partial Response to Intervenor's First Set of Quality Assurance Interrogatories and Request to Produce on October 9, 1985.

4) To the best of my knowledge and belief, the statements contained in the ar.swers to Specific Interrogatories 3, 4 to the extent it relates to 3, and 17 of Applicant's Tenth Partial RvJponse to Intervenor's First Set of Quality Assurance Interrogatories and Request to Produce are true and correct.

Further affiant says not.

)4 , b EM E. Fitzpate Signed and sworn to before me thisM8 day of October 1985 l'Lwk % ((01

~

U Nota h Public

((LL MMhd I b n/:so/cs 0304H .

J

e AFFIDAVIT OF LOUIS 0. DELGEORGE ED Comets.fsPCMDENCE I. Louis O. DelGeorge, being first duly sworn, depose and state as follows:

1) I am employed by comonwealth Edison Company as an Assistant Vice President. 00U'E'EL USNRC

2) My business address is 72 W. Adams St., P.O. Box 767 Chicago, Illinois 60690. l 65 0CT 15 A10:29

3) I am jointly responsible with Mr. Eugene E. Fitzpatrick for the preparation of the introduction to the answer to SpecifleFFuv j g c,, #. j Interrogatories 58 and 59 flied by Intervenors Rorem, et.al^v'fI]hm & SEE - )

also responsible for the portions of the answer to Interrogator {fgC4 J 58 and 59 which respond to the following Contention Items: Item '

1.A; Item 10.A to the extent it crosareferences to Item 1.A; and Item 12.A. This answer was prepared and submitted in Applicant's Sixth Partial Response to Intervenor's First Set of Quality Assurance Interrogatories and Request to Produce on or about August 27, 1985. More recently, I am responsible for the preparation of the answer to Specific Interrogatories 5, 6, 7 and 8 to the extent they relate to 5 and 6, and 57 submitted in Applicant's Tenth Partial Response on October 9, 1985.

4) To the best of my knowledge and belief, the statements contained in the introduction to the answer to Specific Interrogatories 58 and 59, thw portions of the answer which respond to contention Items 1.A, 10.A to the 6xtent it cross-references to 1.A, and 12.A and the answers to specific Interrogatories 5, 6, 7 and 8 to the extent they relate to 5 and 6, and 57 of Applicant's Tenth Partial Response to Intervenor's First Set of Quality Assurance Interrogatories and Request to Produce are true and correct.

5) I also obtained copies of the documents provided in the Supplemental Response to Specific Interrogatory 36 included in Applicant's Seventh Partial Response to Rorem's First Set of Quality Assurance Interrogatories from comonwealth Edison personnel. To the best of my knbwledge and belief, the documents provided describe the programs for field verification of piping components installed at LaSalle anti Byron.

Further affiant says not.

Louis O. DelGeorge 7

[

Signed and sworn to before me this @ day of October 1985 Nf $'

d Notarf Public af lcrut%tQce fJ fO

/Cf30/TS 0305H

s O RELATED CORRESPONDENCE CO:Kr g :-E0

'85 00T 15 A10:29 (f f. [ ' i F El t. -

'. ') '. .

% " A f. (- 4 INDEX OF DOCUMENTS WITHHELD BY APPLICANT UNDER CLAIMS OF PRIVILEGE

TYPE OF BATES PRIVILECE DOCUMENT NUMI*C AC or WP (SO'JRCE) DATE TO FROM DOCUMENT DESCRIPTION 0:001-00008 AC, WP Notebook 04/15/85- R. Lauer's handwritten notes for various (R. Lauer) 07/09/85 conversation with Ceco personnel (Shamblin, Fitzpatclck, Kaushal, Barnes, Schroeder) 00039-00013 AC, WP Draft 11/15/84 Smith Farrar R. Lauer's copy of draft responge to I.R.

Response 84-17 with handwritten coment1 by R. Lauer (R. Lauer)

C0014-00083 AC, WP Report 05/31/85 Report on various Braidwood Corrective (R. Lauer) Action Programs prepared by various CECO personnel at request of and for use by Isham, Lincoln and Beale in anticipation of litigation 00084-00088 WP Handwritten 01/18/85 R. Lauer's notes from 01/18/85 CAT exit Notes (R. Lauer) 00089-00101 WP 'endwritten 05/20/85 R. Lauer's handwritten notes from Keppler/

WMs Warnick desposition taken on 5/20/85 J. Lauer) 4 00102-00105 WP Handwritten 05/23/85 R. Lauer's handwritten from Keppler/

Notes Warnick desposition taken on 5/23/85 (Lauer) 00106-00108 WP Handwritten 02/14/85 R. Lauer's handwritten notes concerning Notes INPO meeting at Region III (Lauer) 00109-00131 WP Handwritten 12/20/83 J. Bloom's handweltten notes from 12/20/83 Notes enforcement conference on I.R. 83-09 (J. Bloom) 1 ____ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

. s TYPE OF BATES P3IVILEGE DOCUMENT NUMBE2 AC or WP (SOURCE) DATE TO FROM DOCUMENT DESCRIPTION 00132-00141 WP Handwritten 03/07/84 J. Bloom's handwritten notes from 03/07/84 Notes enforcement conference related to 83-09 (J. Bloom) 00142 AC, WP Memo 07/03/85 Malman Miller Memo concerning Management Assessment in (Miller) preparation for licensing hearings 00143-00158 AC, WP Memo 07/03/85 Marcus Miller Memo concerning Management Assessment in (Miller) preparation for licensing hearings 00159-00164 AC, WP Memo Memo from Quality First to M. Miller (Miller) concerning interrogatories prepared at request of M. Miller 00165-00170 AC, WP Mem 07/18/85 Wallace Marcus Memo for M. Miller on Open Items issued as (Miller) part of Management Assessment 00171-00177 AC, WP Nemo 07/25/85 Wallace Marcus Memo for M. Miller on Open Items issued as (Miller) part of Management Assessment 00178-00194 AC, WP Report 06/06/85 Marcus, Report for M. Miller concerning (Miller) Lauterbach, Management Assessment Todd 00195 AC, WP Handwritten 06/18/85 M. Miller's notes on MTV and BCAP Notes (Miller) 00196-00200 WP Draft 05/10/85 Smith Farrar M. Miller comments on draft response to Response CAT concerns (Miller) 2

a >

TYPE CF BATES PRIVILEGE DOCUMENT NUMBER AC or WP (SOURCE) DATE TO FROM DOCUMENT DESCRIPTION 00201-00218 WP Draft 05/10/85 Keppler Farrar M. Miller casuments on draft response Response to I.R. 84-44, 84-40 (Miller) 00219 AC, WP Memo 02/03/84 O'Connor Delley Memo re: Braidwood licensing proceeding (Miller) 00220-00221 AC, WP Handwritten 06/18/85 Handwritten notes by Gary C. Jones of S&L Notes about meeting on 6/18/85 concerning MTV (Jones) Attendees: M. Miller, R. Lauer of IL & B; L. DelGeorge, M. Wallace, C. Schroeder, D. Skoza D. Shamblin, B. Shelton, E. Fitzpatrick, W. Vahle, T. Malman, R. Francouer of CECO; T. O'Connor of PGCo 00222 AC, WP Handwritten 06/24/85 Cary C. Jones of S&L handwritten notes of Notes meeting on 6/24/85 concerning MTV.

(Jones) -

Attendees: Ceco personnel and R. Leuer of IL&B 00223-00237 AC, WP Memo 11/12/82 DelCeorgo J. Bloom Legal memo relating to I.R. 82-05 T. Tramm 03238-00242 AC Memo 02/28/84 J. O'Connor D. Swartz Memo discussing I.R. 83-09 B. Thomas C. Reed L. DelGeorge D. Farrar J. Bloom 00243 NOT RESPONSIVE 00244-00250 AC Memo 02/28/84 Same as Same as same as 00238-00242 (Smith) 00238-00242 00238-00242 3

. /

TYPE OF BATES POIVILECE DOCUMENT NUMBER AC or WP (SOURCg) DATE TO FROM DOCUMENT DESCRIPTION 00249-00250 AC, WP Memo 07/18/85 Wallace Marcus Report for M. Miller concerning open items (Marcus) Issued as part of Management Assessment

00251-00266 WP Activity 07/02/85 Prepared as part of Management Assessment Kvaluation for M. Miller (Marcus) 00267-00280 WP Activity 06/18/85 Prepared as part of Management Assessment Evaluation for M. Miller (Marcus) 00281-00291 WP Report 06/06/85 Marcus, Report for M. Miller concerning Marcus (Marcus) Lauterbach, Management Assessment Todd 00292 WP Memo 06/17/85 Description of Management Assessment Program (Marcus) for M. Miller 00293-00294 WP Memo Description of Management Assessment Program (Marcus) 00295-00299 WP Report Prepared in the course of the Management (Marcus) Assessment for M. Miller 00300-00303 WP Memo 06/27/85 Fitzpatrick Marcus Concerns report for M. Miller on Management j (Marcus) Assessment 00304-00307 WP Report S Prepared during the course of the Management (Marcus) Assessment for M. Miller i

00308 WP Handweltten 07/15/85 Prepared during the course of the Management Notes Assessment for M. Miller (Marcus) 4

- a TYPE OF BATES P2IVILEGE DOCUMENT NUMBER AC or WP (SOURCE) DATE TO FROM DOCUMENT DESCRIPTION 0037,9 NOT RESPONSIVE 00310-00318 WP Activity 07/12/85 Prepared during the course of Management Evaluation Assessment for M. Miller (Marcus) 00319 NOT RESPONSIVE' 00320-00329 WP Activity 07/12/85 Prepared during the course of Management Evaluation Assessment for M. Miller (Marcus)

I 00330 NOT RESPONSIVE 00331-00336 WP Activity 07/16/85 Prepared during the course of Management Evaluation Assessment for M. Miller (Marcus) 00337-00345 WP Activity 07/02/85 Prepared during the course of Management Evaluation Assessment for M. Miller (Marcus)

, 00346-00366 WP Notes 06/18/85 Notes concerning Management Assessment for (Marcus) M. Miller 1

00367-00368 WP Notes 07/19/85 Prepared during the course of the Management (Marcus) Assessment for M. Miller 00369-00377 WP Notes 05/24/85 Notes concerning Management Assessment Program j (Marcus) being done for M. Miller i

5 __.--.____

TYPE OF BATES PRIVILEGE DOCUMENT NUMBER AC or WP (SOURCE) DATE TO FROM DOCUMENT DESCRIPTION 00378-00386 WP Notes 05/22/85 Notes concerning Management Assessment Program (Marcus) 00387-00388 WP Notes 05/23/85 Notes concerning Management Assessment Program 00389-00391 WP Notes 05/23/85 Notes concerning Management Assessment Program 00392-00408 WP Notes 04/26/85 Notes concerning Management Assessment Program 00409 WP Notes 05/24/85 Notes concerning Management Assessment Program (Marcus) 00410 WP Notes 05/24/85 Notes concerning Management Assessment Program (Marcus) 00411-00413 WP Notes 05/24/85 Notes concerning Management Aasessment Program (Marcus) l 00414-00420 WP Notes 06/18/85 Notes concerning Management Assessment Program (Marcus) i 00421-00434 AC, WP Report Report on items in contention prepared for use (Schroeder) by CECO management and IL&B

00435-00451 AC, WP Index 06/27/85 Index of Contention Items prepared for use by (Schroeder) Ceco management and IIAB i

i 6

TTPE OF BATES PRIVILEGE DOCUMENT EUMBER AC or WP (SOURCg) DATE TO FROM DOCUMENT DESCRIPTION 00452-00609 AC, WP Report 06/27/85 Report on Contention Items generated for use

(Schroeder) by CRCo management and IL&B 00613-00612 AC, WP Handwritten 07/11/85 R. Lauer's handwritten notes concerning MTV Notes

( m er) 00613-00620 AC, WP Handwritten 07/03/85 R. Lauer's handwritten notes concer'ing MTV Notes (Lauer) 00621-00623 AC, WP Draft Memo M. Miller R. Lauer R. Lauer's draft meno concerning 06/25/85 MTV (Lauer) J. Callo public meeting I

00624-00630 AC, WP Handwritten 06/25/85 R. Lauer's notes on 06/25/85 MTV Public Notes Meeting j (Lauer)

\

) 00631-00640 AC, WP Draft Report 07/15/85 Keppler Ceco Draft 50.55(e) report prepared for 1 (IL&B) attorney review j 00641-00665 AC Draft Report 10/30/84 M. Miller Schroeder Draft Report concerning quality of l (ILAB) R. Lauer Braidwood which was provided for lawyers review I

00666-00716 WP Memo 05/10/85 M. Miller R. Lauer Memo concerning Braidwood Corrective Action (Lauer) J. Callo Program 4

i j 00717-00779 AC, WP Draft 08/28/94 Draft of Braidwood IEPO response with Beeponse handwritten notee by M. Miller (1145)

I 1

) 00780-00802 AC. WP Draft Report Draft Report related to Comstock QC j (IIAs) inspectors allegations of March 29, 1985 i with handwritten notes by R. Lauer

! 7 _ _ _ _ _ _ _ _

-(o TABLE 3.8-2 LIST OF SPECIFICATIONS, CODES, AND STANDARDS

• l SPECIFICATION SPECIFICATION REFERENCE OR STANDARD DESIGNATION TITLE NUMBER 1 ACI 318-71,77 Building Code Requirements for Reinforced Concrete ACI 301 Specifications for Structural 2 Concrete for Buildings 3 ACI 347 Recommended Practice for ANSI A145.1 Concrete Formwork w 4 ACI 305 Recommended Practice for Hot R ANSI A170.1 Weather Concreting y

'co "2 1 ACI 211.1 Recommended Practice for p $

m 5 '

D Selecting Proportions for Normal Weight Concrete 9

6 ACI 304 Recommended Practice for g Measuring, Mixing, Trans- w porting, and placing concrete $

{8 7 ACI 315 Manual of Standard Practice OC 3 for Detailing Reinforced c;

Concrete Structures g

4 8 ACI 306 Recommended Practice for CJ Cold Weather Concreting g

] z UE 0 References to edition dates are shown for the codes used in the design of safety-related $@4 '

structures. All other specifications delineated in this table are recommended practices g and material specifications that do not affect the design of safety-related structures.

• O .

TABLE 3.8-2 (Cont'd)

SPECIFICATION SPECIFICATION REFERENCE OR STANDARD NUMBER DESIGNATION TITLE 9 ACI 309 Recommended Practice for Consolidation of Concrete 10 ACI 308 Recommended Practice for Curing Concrete 11 ACI 214 Recommended Practice for ANSI A146.1 Evaluation of Compression Test Results of Field 12 ACI 311 Recommended Practice for w Concrete Inspection m

, s 13 ACI 304 Preplaced Aggregate Concrete T i for Structural and Mass $

4 Concrete g 14 Report by ACI Placing Concrete by Pumping Committee 304 Method

15 AISC-69,78 Specification for the Design, Fabrication, and Erection of Structural Steel for Building Structural Welding Code 16 AWS Dl.l** -

I et 3 es u go A ** Clarifications'tW and deviations from portions of AWS Dl.1, " Structural Welding to N Nc' Code", are made based on engineering evaluations. 5$

3 w"

• a e .

TABLE 3.8-2 (Cont'd)

SPECIFICATION SPECIFICATION REFERENCE OR STANDARD DESIGNATION TITLE NIMBER ASME Boiler & Pressure Vessel Code, 17 Section III ASME-1971, S73 Division 1, Subsection NE ASME-1974, S75 Division 1, Subsection NF ASME-1973 Division 2, Proposed Standard Code for Concrete Reactor Vessels and Containments Issued for Trial Use and Corments ASME-1980 Division 2, CC 6000

. 18 American Public Test Methods Sulphides in R

• Health Assoc. Water, Standard Methods  ?

• (APHA) for the Examination of M

• Water and Waste Water $

w 19 ASTM Annual Books of ASTM Standards 20 CRSI Manual of Standard MSP-1 Practice 21 ANSI N45.2.5 Proposed Supplementary Q.A. Requirements for Installation, Inspection and Testing of Structural mi Concrete and Structural

$ Steel During Construction g>

fl Phase of Nuclear Power <k Gm Plants r@

if E5

$h "5

0

TABLE 3.8-2 (Cont'd)

SPECIFICATION SPECIFICATION REFERENCE OR STANDARD DESIGNATION TITLE NIMBER 22 CRD Chief of Research and Development Standards, Department of the Army, Handbook for Concrete and Cement Volume I and II, Corps of Engineers U.S. Army ACI-349-76 Code Requirements for 23 Nuclear Safety Related Concrete Structures (D w

• AISI Specification for design of D 24 l cold-formed steel structural 4 e

members y

o l

I m

> M

& 36 xx a

! G til m E a :o d Cn 50 W

B/D-FSAR AMENDMENT 43 SEPTEMBER 1983 4

TABLE 3.8-2 (Cont'd)

EXPLANATION OF ABBREVIATIONS ACI -

Amer ican Concrete Institute AI SC -

American Institute of Steel Construction AISI -

American Iron and Steel Institute ANSI -

American National Standards Institute ASME - American Soclety of Mechanical Engineers ASTM - American Society of Testing Materials AWS - Amer ican Welding Soclety CRD - Chief of Research and Development Standards CRSI - Concrete Reinforcing Steel Institute NOTE: For exceptions and additions to these codes and standards refer to Appendix B.

i 3.8-50 l

LSEJ 100143r6

B/B-FSAR AMENDMENT 22 SEPTEMBER 1979 CHAPTER 17.0 - QUALITY ASSURANCE The quality assurance program for the Byron and Braidwood Stations is conducted in accordance with the Commonwealth Edison Company Quality Assurance Program for Nuclear Generating Stations. This progt;m was initially submitted to the Nuclear Regulatory Commission in June 1975, as Topical Report CE-1. By letter of December 29, 1975, the NRC's Chief of Quality Assurance Branch, Division of Reactor Licensing, informed Commonwealth Edison's Executive Vice President, Mr. W. B. Behnke, that this Topical Report CE-1 was an acceptable program for the design, procurement, con-struction and operations activities within Commonwealth Edison's scope of work for nuclear power plants. Instructions were given to reference this Topical Report in Section 17 of License applications.

Therefore the Commonwealth Edison Topical Report CE-1-A, Revision 7 and all subsequent revisions unless otherwise noted in this chapter, is the basis for the Quality Assurance Program at the Byron /Braidwood Stations.

Section 2.2 of the above referenced topical report provides for compliance with certain regulatory guides and ANSI standards.

Any exceptions taken to the requirements of these guides are found in Appendix A of this FSAR. Otherwise, Byron /Braidwood Stations are in compliance with the guides and ANSI standards.

Subsection 3.2.1.1 of the FSAR states that Safety Category I systems or portions of systems and components meet the require-ments of Appendix B to 10CFR50. Those structures, systems, and components classified as Safety Category I, therefore, are under control of the QA program and are listed in Table 3.2-1. Where only portions of a system are classified as Safety Category I, the P&ID's identify the boundary between Safety Category I and Safety Category II.

l 1

l l

J1 'A00143n?

L .

t. b B/B-FSAR AMENDMENT 18 JANUARY 1979 APPENDIX A - APPLICATION OF NRC REGULATORY GUIDES TABLE OF CONTENTS PAGE A0 INTRODUCTORY MATERIALS A0.0 INDEX TO REGULATORY GUIDES IN TiiIS APPENDIX A0.0-2 AO.1 INTRODUCTION A0.1-1 A1.0 DIVISION 1 REGULATORY GUIDES A1.1-1 A8.0 DIVISION 8 REGULATORY GUIDES A8.8-1 >

i 1

~

} (

i i

l f

3

1 I

l l A0.0-1 06J A00143.s ,

I

B/B-FSAR AMENDMENT 18 JANUARY 1979 A0. 0 - INDEX TO REGULATORY GUIDES IN THIS APPENDIX A1.0 DIVISION 1 REGULATORY GUIDES REGULATORY GUIDE NUMBER TITLE PAGE 1.1 Net Positive Suction Head for Emer-gency Core Cooling and Containment Heat Removal System Pumps A1.1-1 1.2 Thermal Shock to Reactor Pressure Vessels A1.2-1 1.3 Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss-of-Coolant Accident for Boiling Water Reactors A1.3-1 1.4 Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss-of-Coolant Accident for Pressurized Water Reactors A1.4-1 1.5 Assumptions Used for Evaluating the Potential Radiological Consequences of a Steamline Break Accident for Boiling Water Reactors A1.5-1 1.6 Independence Between Redundant Standby (Onsite) Power Sources and Between Their Distribution Systems A1.6-1 1.7 Control of Combustible Gas Concentrations in Containment Following a Loss-of-Cool-ant Accident A1.7-1 1.8 Personnel Selection and Training A1.8-1 1.9 Selection of Diesel-Generator Set Capacity for Standby Power Supplies A1.9-1 1.10 Mechanical (Cadweld) Splices in Rein-forcing Bars of Category I Concrete Structures A1.10-1 1.11 Instrument Lines Penetrating Primary Reactor Containment A1.11-1 1.12 Instrumentation for Earthquakes A1.12-1 1.13 Spent Fuel Storage Facility Design-Basis A1.13-1 1.14 Reactor Coolant Pump Flywheel Integrity 1.15 A1.14-1 Testing of Reinforcing Bars for Cate-gory I Concrete Structures A1.15-1 1.16 Reporting of Operating Infctnation -

Appenc'.ix A Technical Spe<:ifications A1.16-1 1.17 Prctection of Nuclear Power flants Againat Sabotage A1.17-1 1.18 Stru :tural Acceptance Test for Concrete Primary Reactor Cc.ita:nment A1.18-1 i

A0.0-2 A003A3c3 a

r ,

3 5 3/B-FSAR AMENDMENT 18 JANUARY 1979 DIVISION 1 REGULATORY GUIDES (Cont'd)

REGULATORY l GUIDE NUMBER {

TITLE PAGE 1.19 Nondestructive Examination of Primary Containment Liner Welds A1.19-1 1.20 Vibration Measurements on Reactor Internals A1.20-1 1.21 Measuring, Evaluating, and Reporting Radioactivity in Solid Wastes and Releases of Radioactive Materials in Liquid and Gaseous Effluents from Light-Water-Cooled Nuclear Power Plants A1.21-1 1.22 Periodic Testing of Protection System Actuation Functions A1.22-1 1.23 Onsite Meteorological Programs A1.23-1 1.24 Assumptions Used for Evaluating the Potential Radiological Consequences of a Pressurized Water Reactor Radioactive Gas Storage Tank Failure A1.24-1 1.25 Assumptions used for Evaluating the

/

Potential Radiological Consequences of a Fuel Handling Accident in the Fuel Handling and Storage Facility for Boiling and Pressurized Water Reactors A1.25-1 1.26 Quality Group Classifications and Standards for Water , Steam , and Radio-active-Containing Components of Nuclear Power Plants A1.26-1 1.27 Ultimate Heat Sink for Nuclear Power Plants A1.27-1 1.28 Quality Assurance Program Requirements (Design ~and Construction) A1.28-1 1.29 Seicnic Design Classification A1.29-1 1.30 Quality Assurance Requirements for the Installation, Inspection, and Testing of Instrumentation and Electric Equip-ment A1.30-1 1.31 Control of Stainless Steel Welding A1.31-1 1.32 Use of IEEE Standard 308-1971, " Criteria for Class IE Electric Systems for Nu-clear Power Generating Stations" A1.32-1 1.33 Quality Assurance Program Requirements (Operation) A1.33-1 1.34 Control of Electroslag Weld Properties A1.34-1 1.35 Inservice Inspection of Ungrouted Ten-dons in Prestressed Concrete Contain-ment Structures A1.35-1 A0.0-3 10014370

~

. $ B/B-FSAR AMENDMENT 18 JANUARY 1979 i

i l ,

a

_Df_YISION 1 REGULATORY GUIDES (Cont'd) I l

I i

REGULATORY 1

GUIDE NUMBER TITLE PAGE j 1.36 Nonmetallic Thermal Insulation for i Austenitic Stainless Steel A1.36-1

! 1.37 Quality Assurance Requirements for

] Cleaning of Fluid Systems and Associ-I ated Components of Water-Cooled Nuclear .

j Power Plants A1.37-1 l 1.38 Quality Assurance Requirements for i Packaging, Shipping, Receiving, Storage, j and Handling of Items for Water-Cooled Nuclear Power Plants A1.38-1

! 1.39 Housekeeping Requirements for Water-1 Cooled Nuclear Power Plants A1.39-1

, 1.40 Qualification Tests of continuous-

Duty Motors Installed Inside the Con-

) tainment of Water-Cooled Nuclear Power j Plants A1.40-1 1.41 Preoperational Testing of Redundant j Onsite Electric Power Systems to Verify Proper Load Group Assignment A1.41-1

( 1.42 Interim Licensing Policy on As Low As Practicable for Gaseous Radiciodine from Light Water-Cooled Nuclear Power Reactors A1.42-1 4

1.43 Control of Stainless Steel Weld Clad-i ding of Low-Alloy Steel Components A1.43-1 1.44 Control of the Use of Sensitized Stain-

less Steel A1.44-1 J 1.45 Reactor Coolant Pressure Boundary Leak-age Detection Systems A1.45-1 l 1.46 Protection Against Pipe Whip Inside t Containment A1.46-1

! 1.47 Bypassed and Inoperable Status Indica-

! tion for Nuclear Power Plant Safety Systems A1.47-1 I 1.48 Design Limits and Loading Cominations for Seismic Category I Fluid System i Components A1.48-1 i 1.49 Power Levels of Nuclear Power Plants A1.49-1 l 1.50 control of Preheat Temperature for i Welding of Low-Alloy Steel A1.50-1 j 1.51 Inservice Inspection of ASME Code Class j 2 and 3 Nuclear Power Plant Components A1.51-1 l 1.52 Design, Testing, and Maintenance Criteria

! for Atmosphere cleanup System Air Filtra-l tion and Adsorption Units of Light-Water-l Cooled Nuclear Power Plants A1.52-1 r

A0.0-4 a

__ _ _ _ _ _ _ _ _ ~ _ . - - _ _ _ = _ _ _ . _ _ . . _ . _--

, s B/B-FSAR AMENDMENT 20 MAY 1979 i

a

, DIVISION 1 REGULATORY GUIDES (Cont'd) i REGULATORY

GUIDE NUMBER TITLE PAGE 1.53 Application of the Single-Failure Cri-terion to Nuclear Power Plant Protection Systems A1.53-1

1.54 Quality Assurance Requirements for Pro-i tective Coatings Applied to Water-Cooled Nuclear Power Plants A1.54-1 1.55 Concrete Placement in Category I Struc-tures A1.55-1 1.56 Maintenance of Water Purity in Boiling Water Reactors A1.56-1

, 1.57 Design Limits and Loading Combinations t for Metal Primary Reactor Containment System Components A1.57-1 1.58 Qualification of Nuclear Power Plant i Inspection, Examination, and Testing Personnel A1.58-1 1.59 Design-Basis Floods for Nuclear Power Plants A1.59-1 1.60 Design Response Spectra for Seismic I Design of Nuclear Power Plants A1.60-1 I 1.61 Damping Values for Seismic Design of Nuclear Power Plants A1.61-1

1.62 Manual Initiation of Protective Actions A1.62-1 1.63 Electric Penetration Assemblies in Containment Structures for Water-Cooled i Nuclear Power Plants. A1.63-1 l 1.64 Quality Assurance Requirements for the

] Design of Nuclear Power Plants A1.64-1 1.65 Materials and Inspection for Reactor Vessel Closure Studs A1.65-1

1.66 Nondestructive Examination of Tubular

, Products A1.66-1 1.67 Installation of Overpressure Protec-tion Devices A1.67-1 1.68 Preoperational and Initial Startup Test Programs for Water-Cooled Power Reactors A1.68-1

1.68.1 Preoperational and Initial Startup i Testing of Feedwater and Condensate Systems for Boiling Water Reactor

, Power Plants A1.68.1-1

1.68.2 Initial Startup Test Program to

! Demonstrate Remote Shutdown Capabi.'6y for Water-Cooled Nuclear Power Plants A1.68.2-1 i

A003A372 A0.0-5 l

l

. s C/D-FSAR AMENDMENT 18 JANUARY 1979 DIVISION 1 REGULATORY GUIDES (Cont'd)

REGULATORY GUIDE NUMBER TITLE PAGE 1.69 Concrete Radiation Shields for Nuclear Pouer Plants A1.69-1 1.70 Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants A1.70-1 1.71 Welder Qualification for Areas of Limited Accessibility A1.71-1 1.72 Spray Pond Plastic Piping A1.72-1 1.73 Qualification Test of Electric Valve Operators Installed Inside the Contain-ment of Nuclear Power Plants A1.73-1 1.74 Quality Assurance Terms and Definitions A1.74-1 1.75 Physical Independence of Electric Sys-tems A1.75-1 1.76 Design-Basis Tornado for Nuclear Power Plants A1.76-1 1.77 Assumptions Used for Evaluating a Control Rod Ejection Accident for Pressurized Water Reactors A1.77-1

( 1.78 Assumptions for Evaluating the Habit-ability of a Nuclear Power Plant Control Room During a Postulated Hazardous Chemical Release A1.78-1 1.79 Preoperational Testing of Emergency Core Cooling Systems for Pressurized Water Reactors A1.79-1 1.80 Preoperational Testing of Instrument

. Air Systems A1.80-1 1.81 Oshared Emergency and Shutdown Electric Systems for Multiunit Nuclear Power Plants A1.81-1 1.82 Sumps for Emergency Core Cooling and Containment Spray Systems A1.82-1 1.83 Inservice Inspection of Pressurized Water Reactor Steam Generator Tubes A1.83-1 1.84 Code case Acceptability-ASME Section III Design and Fabrication A1.84-1 1.85 Code caso Acceptability-ASME Section III Materials A1.85-1 1.86 Termination of Operating Licenses for Nuclear Reactors A1.86-1 1.87 Guidance for Construction of Class I Components in Elevated-Temperature Reactors A1.87-1 l

l l A0.0-6 A')0113'3

B/B-FSAR AMENDMENT 18 JANUARY 1979 DIVISION 1 REGULATORY GUIDES (Cont'd)

REGULATORY GUIDE NUMBER TITLE PAGE 1.88 Collection, Storage, and Maintenance of Nuclear Power Plant Quality Assurance Records A1.88-1 1.89 Qualification of Class 1E Equipment for Nuclear Power Plants A1.89-4 1.90 Inservice Inspection of Prestressed Concrete Containment Structures with

, Grouted Tendons A1.90-1

1. 9'1 Evaluation of Explosions Postulated to Occur on Transportation Routes Near Nuclear Power Plants A1.91-1 1.92 Combining Modal Responses and Spacial Components in Seismic Response Analysis A1.92-1 1.93 Availability of Electric Power Sources A1.93-1 1.94 Quality Assurance Requirements for In-stallation, Inspection and Testing of Structural Concrete and Structural Steel During the Construction Phase

( of Nuclear Power Plants A1.94-1 1.95 Protection of Nuclear Power Plant Control Room Operators Against an Accidental Chlorine Release- A1.95-1 1.96 Design of Main Steam Isolation valve Leakage Control Systems for Boiling Water Reactor Nuclear Power Plants A1.96-1 1.97 Instrumentation for Light-Water-cooled Nuclear Power Plants to Assess Plant Conditions During and Following an Accident A1.97-1 1.98 Assumptions Used for Evaluating the Potential Radiological Consequances of a Radioactive Offgas System Failure in a Boiling Water Reactor A1.98-1 1.99 Effects of Residual Elements or Pre-dicted Damage to Reactor Vessel Materials A1.99-1 1.100 Seismic Qualification of Electric Equipment for Nuclear Power Plants A1.100-1 1.101 Emercency Planning for Nuclear Power Plants A1.101-1 1.102 Flood Protection for Nuclear Power Plants A1.102-1 1.103 Post-Tensioned Prestressing Systems for Concrete Reactor Vessels and Containments A1.103-1 A0014374 A0.0-7 r-

AMENDMENT 43 B/B-FSAR SEPTEMBER 1983 DIVISION 1 REGULATORY GUIDES (Cont'd)

REGULATORY GUIDE PAGE NUMBER TITLE 1.104 Overhead Crane Handling Systems for Nuclear Power Plants A1.104-1 1.105 Instrument Setpoints Al.105-1 l 1.106 Thermal Overload Protection for Electric A1.106-1 Motors on Motor-Operated Valves 1.107 Qualification for Cement Grouting for Prestressing Tendons in Containment Structures A1.107-1 1.108 Periodic Testing of Diesel Generators Used as Onsite Electric Power Systems at Nuclear Power Plants A1.108-1 1.109 Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I A1.109-1 1.110 Cost Benefit Analysis for Radwaste Systems for Light-Water-Cooled Nuclear Power Reactors A1.110-1 1.111 Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors A1.111-1 1.112 Calculation of Releases of Radio-active Materials in Gaseous and Liquid Effluents from Light-Water-Cooled Power Reactors A1.112-1 1.113 Estimating Aquatic Dispersion of Effluents from Accidental and Routine Reactor Releases for the Purpose of Implementing Appendix I A1.113-1 1.114 Guidance on Being Operator at the Controls of a Nuclear Power Plant A1.114-1 1.115 Protection Against Low-Trajectory Turbine Missiles A1.115-1 1.116 Quality Assurance Requirements for Installation, Inspection and Testing of Mechanical Equipment and Systems A1.116-1 1.117 Tornado Design Classification A1.117-1 1.118 Periodic Testing of Electric Power and Protection Systems A1.118-1 1.119 Surveillance Program for New Fuel Assembly Designs A1.119-1 1.120 Fire Protection Guidelines for Nuclear Power Plants A1.120-1 1.121 Bases for Plugging Degraded PWR Steam Generator Tubes A1.121-1 l alm!14375 A0.0-8 l

l l

B/B-FSAR AMENDMENT 43 SEPTEMBER 1983 LIVISION 1 REGULATORY CUIDES (Cor.t ' d )

REGULATORY CUILE TITLE PAGE NUMBER 1.122 Development of Floor Design 1:esponse Spectra for Seismic Design of Floor-Suppcrted Equipment or Cenponents Al.122-1 1.123 Cuality Ascurance Requirements for Control of Procurement of Items and Services for Nuclear Power Plants A1.123-1 1.124 Service LLmits and Loading Continations for Class I Linear-Type Suppcrts A3.124-1 1.125 Physical Models for Design and Operation of flydraulic Structures and Systems for Nuclear Power Plants A1.125-1 1.126 An Acceptable !!cdel and Related Statistical Methods for the 1.nalysis of Fuel Densificaticn Al.126-1 1.127 Inspection of Water Control Structures Associated with Nuclear Power Plants A1.127-1 1.128 Installation Design and Installation of Large Lead Storage Batteries for Nuclear Power Plants A1.126-1 1.129 Maintenance, Testing, and Replacement of Large Lead Storage Batteries for Nuclear Fower Plants A1.129-1 1.130 Design Limits and Loading Combinations for Class I Plate-and-Shell-Type Component Supports A1.130-1 1.131 Qualification Tests of Electric Cables, Field Splices and Connections for Light-Water-Cooled Nuclear Power Reactors A1.131-1 1.132 Site Investigations for Foundations of Nuclear Power Plants A1.132-1 1.133 Loose Part Detection Program For the Primary System of Light Water-Cooled Rcactors A1.133-1 1.134 Medical Certification and Monitoring of Personnel Requiring Operator. Licenses A1.134-1 1.135 Normal Water Level and Discharge at Nuclear Power Plants A1.135-1 1.136 Material for Concrete Containments A1.136-1 1.137 Fuel-Oil Systems for Standby Diesel Generators A1.137-1 1.138 Laboratory Investigations of Soils for Engineering Analysis and Design of Nuclear Power Plants 11.138-1 1.139 Guidance for Residual Heat Removal A1.139-1 1.140 Design, Testing and Maintenance Criteria for Normal Ventilation Exhaust System Air Filtration and Adsorption Units of Light-Water-Cooled Nuclear Power Plants Al.140-1 1.141 Ccntainment Isolation Provisions for Fluid Systems A1.141-1 A0.0-9 N7@7G m

, . i AMENDMENT 43 B/B-FSAR SEPTEMBER 1983 i

DIVISION 1 REGULATORY GUIDES (Cont'd)

REGULATORY ,

GUIDE PAGE NUf1BER TITLE 1.142 Safety-Related Concrete Structures for Nuclear Power Plants (Other than Reactor A1.142-1 Vessels and Containments) 1.143 Design Guidance for Radioactive Waste

?

P.anagement Systems, Structures and Components Installed in Light-Water- A1.143-1 Cooled Nuclear Power Plants 1.144 Auditing of Quality Assurance Programs A1.144-1 for Nuclear Power Plants 1.146 Qualification of Quality Assurance Program Audit Personnel for Nuclear A1.146- 1 Power Plants 1

j i

i I

1 k

A0014377 l

l A0.0-9a l - -. - -

B/B-FSAR RMENDMENT 22 SEPTEMBER 1979 A8.0 DIVISION 8 REGULATORY GUIDES REGULATORY

! GUIDE l NUMBER TITLE PAGE i 8.2 Guide for Administrative Practices in Radiation Monitoring A8.2-1 8.3 Film Badge Performance Criteria A8.3-1 8.7 Occupational Radiation Exposure Records System A8.7-1 8.8 Information Relevant to Ensuring i that Occupational Radiation Exposures at Nuclear Power Stations will be as Low as is Reasonably Achievable A8.8-1 i

8.9 Acceptable Concepts, Models, Equations, and Assumptions for a Bioassay Program A8.9-1 1 8.10 Operating Philosophy for Maintaining '

Occupational Radiation Exposures t as Low as is Reasonably Achievable A8.10-1

, 8.12 Criticality Accident Alarm Systems A8.12-1

8.15 Acceptable Programs for Respiratory Protection A8.15-1 8.19 Occupational Radiation Dose Assessment in Light-Water Reactor j

Power Plants Design Stage Man-Rem Estimates A8.19-1 I

i 1

1 i

A00143'78 A-0.0-10

e

• B/B-FSAR AMENDMENT 18 JANUARY 1979 APPENDIX A - APPLICATION OF NRC REGULATORY GUIDES A

0.1 INTRODUCTION

This appendix indicates compliance with NRC Regulatory Guides and

. indicates the parts of the FSAR in which the requirements of the guides are addressed.

1 4

(:

i I

I i ,

I A0034373 A0.1-1 l

I B/B-FSAR AMENDMENT 37 MARCH 1982 REGULATORY GUIDE 1.1 Revision 0, November 2, 1970 NET POSITIVE SUCTION HEAD FOR EMERGENCY CORE COOLING AND CONTAINMENT HEAT REMOVAL SYSTEM PUMPS The Applicant meets all objectives set forth in this Regulatory Guide as presented in Ctbsection 6.3.2.2.

1 k

i i

i i

i A1.1-1 d@7.4310

! i r

I l

B/B-FSAR AMENDMENT 37 MARCH 1982 REGULATORY GUIDE _1. 2 i

! l f i Revision 0, November 2, 1970 l

THERMAL SHOCK TO REACTOR PRESSUFE VtSSELS i

Westingnouse follows all recommendations of the guide. The guide

Position C.1 is followed by Westinghouse's own analytical and

! experimental programs as well as oy participation in the Heavy i Section Steel Technology (HSST) Program at Oak Rioge National Labora tory.

q Analytical techniques have been developed oy Westinghouse to perrorm fracture evaluations of reactor ves.sels unoer thermal snock loadings.

i i

Under the Heavy Section Steel Technology Program a numoer of 6-inch thick 39-inch OD steel pressure vessels containing l carefully prepared and sharpened surf ace cracxs are being tested.

Test conditions include both hyaraulic internal pressure loadir.gs and thermal shock loadings. The objective of this program is to validate analytical fracture mechanics techniques and demonstrate quantitatively the margin of safety inherent in reactor pressure vessels.

A number of vessels have been tested under nydraulic pressure loadings, and results have confirmed the validity of f racture analysis techniques. The results anc implications of the hydraulic pressure tests are summarized in Oak Ridge National j Lanoratory re eort ORNL-TM-5909.

Three thermal shock experiments have oeen completed and are now i

Deing evaluated. Preliminary information indicates that the analytical techniques do agree tavorably with experimental i results.

! Westinghouse is continuing to obtain f racture toughness data for reactor pressure vessel steels through internally f unoed prograns I

as well as HSST sponsored work.

j Fracture toughness testing of irradiated compact tension tracture l toughness specimens has been completed. The complete l postirradiation data on 0.394-inch, 2-inch, and 4-inch thick

,' specimens are now available from the HSSI program. Both static ,

and dynamic postirradiation fracture toaghness data have oeer.

obtained. Evaluation of the data ootained to date on material irradiated to fluences between 2.2 and 4.5 10a ' n/cma indicated A1.2-1 A0014381

B/D-FSAR AMENDMENT 38

' MAY 1982 i

that the reference toughness curve as contained in the ASML Section III code remains a conservative lower bound for toughness values for pressure vessel steels. Another f racture toughness program is now underway. This program involves the irradiation and testing of weld metal used in fabrication of operating pressure vessels of pre-1972 construction. These welds characteristically have high copper contents and low initial charpy '/-Notch " Shelf" energies. Results of this program are ex eected in 1978.

Details of progress and results obtained in the HSST program are available in the Heavy Section Steel Technology Program Semiannual (quarterly beginning in 1974) Progress Reports, issued by Oak Ridge National Laboratory.

Regulatory Position c.2 is followed inasmuch as no significant changes have oeen made in approved core or reactor designs.

The guide position C. 3 is followed since the vessel design does not include the use of an engineering solution to assure adequate recovely of the fracture toughness properties of the vessel material. If additional margin is needed, the reactor vessel can be annealed at any point in its service life. This solution is already feasible, in principle, and could be performed witn the vessel in place.

l A1.2-2 A0014362

i f

i i b/b-1 SAP A!!END!!CNT 37 '

'tARCil 1982 -

Er.GULATOhY GU_10E 1.3 4

Revision 2, June 1974 e

t l

AbdUMPA IONS USED FOR LVALUA.1NG THE Po1EN'I AAL  !

RADIOLOGICAL CONSEQUENCES OF A LOSo-OF-COOLANI ACCIDENI FOR BOILING 4AI Lk hEAC"40hS 4his guide is pertinent to BWF's only.

i 1

l 1

i l

1 1

1 A

A9014383 A1.3-1

- -- -. -.w - . - r-. . ,.= + w, ,r-,w, w e=.ree,. ,g-w , - - - . , ,.e,.

B/B-FSAR AMENDMENT 37 MARCH 19 82 REGULATORY GUIDE 1.4 Revision 2, June 1974 ASSUMPTIONS USED FOR EVALUATING ThE FOTENTI AL RADIOLOGICAL CONSEQUENCES OF A LOSS-OF-COOLANT ,

ACCIDENT FOR PRESSURIZED WATER REACTORS The requirements of this guide have been adhered to in all pertinent sections of this application. The meteorology assumptions from the guide are detailed in Subsection 2.3. 3. The guide assumptions on radioisotope releases are detailed in Section 15.4, as are the assumptions on containment spray effectiveness.

I l

A00143S4 A1.4-1

B/B-FSAR A!!ENDMENT 37 MARCH 19 82 REGULATOEY GUIDE 1.5 Revision 0, March 10, 1971 RdSUMPIIONS USED FOR z.V ALUATING THE POTENTIAL RADIOLOGICAL CONSEQUENCES OF A_STEAMLINE BFi AK ACCIDENT FOR BOILING 4ATER REACTORS ibis guide is pertinent to BWF's only.

ADe14385 A1.5-1

B/B-FSAR AMENDMENT 37 MARCl! 19 82 i

REGULATORY GUIDE 1.6 Revision 0, March 10, 1971 INDEPENDENCE BETWEEN REDUNDANT STANDBY (ONSITE) POWER SOURCES AND BETWEEN THEIR DISTRIBUTION SYSTEMS 4

I The Applicant complies with this Regulatory Guide. Refer to FSAR Subsections 8.1.1 and 8.1.6 for further information.

5 A0014386 A1.6-1

l B/B-FSAR AMENDMENT 37 MARCH 1982  ;

REGULATORY GUIDE 1. 7 Revision 1, September 1976 CONTROL OF COMBUSTIBLE GAS CONCENTRATIONS IN CONTAINMENT FOLLOWING A LOSS-OF-COOLANT ACCIDENT The Applicant complies with the regulatory positions stated in this Regulatory Guide. Refer to Subsections 6.2.5.1 and 6.2.5.2 for further information.

l l

I l

t l

A1.7-1 A;}P'J. '23S7

B/B-FSAR AMENDMENT 37 MARCH 1982 j REGULATORY GUIDE 1.8 Revision 1, September 1975 PERSONNEL SELECTION AND TRAINING The Applicant complies with this Regulatory Guide. Refer to Section 13.1 and Subsection 13.2.4 of the FSAR for further information.

l ADe3.4388 A1.8-1

.a

B/B-FSAR AMENDMENT 37 MARCH 1982 REGULATORY GUIDE 1.9 Revision 2, December 1979 SELECTION, DESIGN, AND QUALIFICATION OF DIESEL-GENERATOR UNITS USED AS ONSITE ELECTRIC POWER SYSTEMS AT NUCLEAR POWER PLANTS The Applicant complies with the regulatory position with th'e following clarification regarding paragraph C.4:

Due to high transformer inrush current, the voltage may dip below the required limit of 75% of nominal upon energiz-ing the 480-Vbit substation transformers and their auxiliary loads. However, this dip is of a very short duration (0.2 to 0.5 seconds) and will occur immediately after the diesel generator breaker is closed. Since the diesel breaker is expected to close 8.5 to 9 seconds following a loss of offsite power (LOOP) and the first motor load (Centr if ugal Charging Pump motor) is sequenced on 10 seconds after a LOOP, the voltage will have recovered to the required limits prior to beginning the load sequence.

Compliance with the requirements of this guide is described further in Subsections 8.1.2, 8.1.20, 8.3.1.1.1 and 8.3.1.2.

Therefore, the Applicant meets the objectives set forth in this Regulatory Guide.

l l

l l

l l

A1.9-1 l AW143SS l w-r

B/B-FSAR AMENDMENT 37 MARCH 1982 REGULATORY GUIDE 1.10 Revision 1, January 2, 1973 MECHANICAL (CADWELD) SPLICES IN REINFORCING BARS OF CATEGORY I CONCRETE STRUCTURES This regulatory guide was withdrawn July 8, 1981, however, the plant design conforms to the regulatory positions as de-scribed in Subsection B.2.3 of Appendix B.

A1.10-1 AT 14390 l

l

B/B-FSAR AMENDMENT 37 MARCH 1982 REGULATORY GUIDE 1.11 Revision 0, March 10, 1971 INSTRUMENT LINES PENETRATING PRIMARY REACTOR CONTAINMENT The Applicant complies with the requirements of this Regulatory Guide as discussed in Subsections 7.1.2.5 and 7.3.1.1.2.

I I

gy,77,7 W.W.M i

I

B/B-FSAR AMENDMENT 37 MARCH 1982 .-

REGULATORY GUIDE 1.12 Revision 1, April 1974 INSTRUMENTATION FOR EARTHQUAKES The plant design conforms to the regulatory positions as dis-cussed in Subsections 3.7.1.1 and 3.7.4.2.

i l

A1.12-1 A0014392

l B/B-FSAR AMENDMENT 37 i MARCH 1982 l l

REGULATORY GUIDE 1.13 Revision 1, December 1975 SPENT FUEL STORAGE FACILITY DESIGN-BASIS The plant design conforms with the requirements of this guide as presented in Subsections 9.1.2.3 and 9.1.3.3.

1 Al.13-1 l AOP14393

B/B-FSAR AMENDMENT 37 MARCH 1982 REGULATORY GUIDE 1.14 o

Revision 1, August 1975 REACTOR COOLANT PUMP FLYWriEEL INTEGRITY The design meets the requirements of Regulatory Guide 1.14 with the exceptions noted below. The shaft and the bearings supporting the flywheel are capable of withstanding any combination of the normal operating loads, anticipated transients, the design-basis LOCA, and the safe shutdown earthquake loads.

Since the issuance of Regulatory Guide 1.14, Revision 1, the NRC Staf t has provided to Westinghouse a copy of Draft 2, Revision 2, of Regulatory Guide 1.14 (via an April 12, 1976, letter from Robert B. Minogue to C. Eiche1dinger) . Tnis draft was tormulated f rom industry and concerned parties' comments. It is significant that the Draft 2 version incorporates several of the Westinghouse comments on Revision 1. Since Draft 2 has not been formally published as Revision 2 of Regulatory Guide 1.14, the exceptions and clarifications (from the original Westinghouse comments) are provided in the following:

a. Post-Spin Inspection Westinghouse has shown in WCAP-8163, " Topical Re e ort on Reactor Coolant Pump Integrity it. LOCA," that the flywheel would not fail at 290% of normal speed for a flywheel flaw of 1.15 inches or less in length.

l Results for a double ended guillotine break at the gump discharge with tull separation or pipe ends assumed, show the maximum overspeed was calculated in WCAP- 8163 to be about 280% of normal speed for the same postulated break, and an assumed instantaneous loss of poder to the reactor coolant pump. In comparison with the overspeed presented above, the flywheel could withstand a speed up to 2.3 times greatcr than the flywheel spin test speed of 125%

provided that flaws no greater than 1.15 inches are present. If the maximum speed were 125% of normal speed or less, the critical flaw size for f ailure would exceed 6 inches in length. Nondestructive tests and critical dimension examinations are all performed before the spin tests. The inspection methods employed (cescribed in WCAP-8163) provide assurance that flaws significantly smaller than the critical flaw size of 1.15 inches for 2904 of normal 1 A1.14-1 A00M3S4

B/B-FSAR speed would be detected. Flaws in the flywheel will be recorded in the prespin inspection program (s ee WCAP- 8163) . Flaw growth attributable to the spin test (i.e., from a single reversal of stress, up to speed and back) , under the most adverse conditions, is about three orders of magnitude smaller than that nondestructive inspection techniques are capable of detecting. For these reasons, Westinguouse does not perform postspin inspections and believes the prespin test inspections are adequate.

b. Interference Fit Stresses and Excessive Deformation Much of Revision 1 deals with stresses in the flywheel resulting from the interference fit between the flywheel and the shaft. Because Westinghouse's design specifies a light interference fit between the flywheel and the shaft; at zero speed, the hoop stresses and radial stresses at the flywheel bore are negligible. Centering of the flywheel relative to the shaft is accomplished by means of keys and/or centering devices attached to the shaft, and at normal speed, the flywheel is not in contact with the shaft in the sense intended by Revision 1. Hence, the definition of " Excessive Deformation," as defined

in Revision 1 of Regulatory Guide 1.14, is not applicable to the Westinghouse design since the enlargement of the bore and subsequent partial separation of the flywheel trom the shaft does not cause unbalance of the flywheel. Extensive Westinghouse experience with reactor coolant pump flywheels installed in this fashion has verified the adequacy of the design.

Westinghouse's position is that combined primary stress levels, as defined in Revision 0 of Regulatory Guide 14 (C. 2 ) , (a) and (c) , are both conservative and proven and that no changes to these stress levels are necessary. Westinghouse designs to these stress limits and thus does not have permanent distortion of the flywheel bore at nornal or spin test conditions.

c. Section B, Discu'ssion of Cross Rolling Ratio of 1 to 3 i

Cross Rolling Ratio - Westinghouse's position is that specification of a cross rolling ratio is unnecessary since past evaluations have shown that ASME SA-533-B Class 1 materials produced without this requirement have suitaole toughness for typical flywheel applications. Proper material selection and

! specification of minimum material properties in the l transverse direction adequately ensure flywheel l

integrity. An attempt to gain isotropy in the A1.14- 2 l Acog395 l _ - . ..

B/B-FSAR AMENDMENT 37

?1 ARCH 19 82 flywheel material by means of cross rolling is unnecessary since adequate margins of safety are provided by both flywheel material selection (ASME SA-533-B Class 1) and by specifying minimum yield and I tensile levels and toughness test values taken in the  :

direction perpendicular to the maximum working direction of the material. I

d. Section C, Item 1a Relative to Vacuum-melting and Degassing Process or the Electroslag Process Vacuum Treatment - The requirements for vacuum melting and degassing process or the electroslag process are not essential in meeting the balance of the Regulatory Position nor do they, in themselves, ensure compliance with the overall Regulatory Position. The initial Safety Guide 14 stated that the " flywheel material should be produced by a process that minimized flaws in the material and improves its fracture toughness properties." This is accomplished by using SA-533 material including vacuum treatment.

e. Section C, Item 2b: Westinghouse interprets this paragraph to mean:

Design Speed Definition - Design speed should be 125%

of normal speed or the speed to which the pump motor might be electrically driven by station turbine generator during anticipated transients, whichever is greater. Normal speed is defined as the synchronous speed of the a-c drive motor at 60 Hz. l The flywheel integrity is described in Subsections 3.8.3.4.1 (3.8.3.2.1), 5.4.1.5.2, and 5.4.1.5.3.

A1.14-3 APD343S6

B/D-FSAR AMENDMENT 37 MARCH 1982 REGULATORY GUIDE 1,15 Revision 1, December 28, 1972 TESTING OF REINFORCING BARS FOR CATEGORY I CONCRETE STRUCTURES The plant design conforms to the regulatory positions as dis-cussed in Subsection 3.8.3.6, Table 3.8-2, and Section B.2 of Appendix B.

i A1.15-1 l

A0034397

B/B-FSAR AMENDMENT 37 MARCH 1982 REGULATORY GUIDE 1.16 Revision 4, August 1975 REPORTING OF OPERATING INFORMATION - APPENDEX A TECHNICAL SPECIFICATIONS The Applicant complies with this Regulatory Guide. Refer to FSAR Chapter 16.0 for further information.

i l

l A1.16-1 -

(105.43S8

B/B-FSAR AME24DMENT 37 MARCH 1982 ,

REGULATORY GUIDE 1.17 Revision 1, June 1973 PROTECTION OF NUCLEAR POWER PLANTS AGAINST INDUSTRIAL SABOTAGE The Industrial Security Plan for the Byron and Braidwood Stations has been provided to the Regulatory Staff on a pro-prietary basis. No comparison of the Plan and this Regulatory Guide will be provided in this response. Compliance with the requirements of this guide is presented in Section 13.6 and Subsection 9.5.2.2.

I i

A0014399 A1.17-1

B/B-FSAR AMENDMENT 37 MARCH 1982 i

l REGULATORY GUIDE 1.18 No Current Issue STRUCTURAL ACCEPTANCE TEST FOR CONCRETE PRIMARY REACTOR CONTAINMENTS This Regulatory Guide was withdrawn July 8, 1981. The structural acceptance test shall conform to the requirements of the ASME Code Section III, Division 2/ACI 359-80 Article CC-6000, as stated in FSAR Subsection 3.8.1.7.2.2 l

l r

l l

A1.18-1 A00M400

B/B-FSAR AMENDMENT 37 MARCH 1982 REGULATORY GUIDE 1.19 Revision 1, August 11, 1972 4

NONDESTRUCTIVE EXAMINATION OF PRIMARY CONTAINMENT LINER % ELDS

'1 This regulatory guide was withdrawn July 8, 1981, however, the plant design conforms to the regulatory positions as de-scribed in Subsections 3.8.1 and 3.8.2.

1 i

A1.19-1 A 001rA01

B/B-FSAR AMENDMENT 37 MARCH 1982 REGULATORY GUIDE 1.20 Revision 2, May 1976 COMPREHENSIVE VIBRATION ASSESSMENT PROGRAM FOR REACTOR INTERNALS DURING PREOPERATIONAL AND INITI AL STARTUP TESTING The requirements of Regulatory Guide 1.20 are met. Refer to Subsection 3.9.2.4 for further discussion.

i i

t A1.20-1 A003.4402 l- -- .

B/B-FSAR AMENDMENT 37 MARCH 19 82 REGULATORY GUIDE 1.21 r

Revision 1, June 1974 MEASURING, EVALUATING, AND REPORTING RADIOACTIVITY IN SOLID WASTES AND RELEASES OF RADIOACTIVE MATERIALS IN LIQUID AND GASEOUS EFFLUENTS FROM LIGHT-WATER-COOLED NUCLEAR POWER PLANTS The Applicant complies with this Regulatory Guide with the following clarifications keyed to paragraph numbers in the Regulatory Position.

10. If multiple sample points are given for detection of radioiodine and readings are below the threshold of detection, the threshold limits will not be summed over the number of sample points to give the total release rate.

14. Sensitivities in Appendixes A and B of this guide may not be practicable. These releases will be measured to the lowest levels consistent with existing technology.

Assurance of measuring very low levels of radioactivity is subject to interpretation of readout which may be effected by noise level, calibration, radiation, background, etc. In j addition, the instrument sensitivity required to assure

' compliance with the guide may not be available with current technology.

i ATM403 A1.21-1 l

B/B-FSAR AMENDMENT 37 MARCH 19 82 1

REGULATORY GUIDE 1.22 Revision 0, February 17, 1972 PERIODIC TESTING OF PROTECTION SYSTEM ACTUATION FUNCTIONS The Applicant complies with this Regulatory Guide. Refer to FSAR Subsections 7.1.2.6, 7.3.2.2, 8.1.3, 8.3.1.2 and 12.3.4.1 for further information.

4 l

l l

l l

l l

l I A1.22-1 l

' A0014404

. o i

l B/B-FSAR AMENDMENT 37 MARCH 19 82 REGULATORY GUIDE 1.23 4

Revision 0, February 17, 1972 ONSITE METEOROLOGICAL PROGRAMS The Applicant complies with this Regulatory Guide. Refer to Subsection 2.3.3 for further information.

i f

I l -

l i

l A1.23-1 i

l A00144G5

. . 1 B/B-FSAR RMENDMENT 37 i tiARCH 1982 (

REGULATOFi GUIDE 1.24 Revision 0, March 23, 1972 i

ASSUMPTIONS USED FOR EVALUATING THE POTENTIAL RADIOLOGICAL CONSEQUENCES OF A PRESSURIZdD WATER REACTOR RADIOACTIVE GAS STORAGE TANK FAILURE Applicant complies with the regulatory position of the guide as presented in Suosection 15.7.1. 3.

l f

A1.24-1 A00244G6

B/B-FSAP A!!ENDf1ENT 37 tiARCH 1982 )

\

REGULATORY GUIDE 1.25 j Revision 0, March 23, 1972 ASSUMPTIONS USED FOR EVALUATING THE POTENTIAL RADIOLOGICAL CONSEQUENCES OF A FUE.L HANDLING ACCIDENT IN THE FUEL HANDLING AND STORAGE FACILITY FOR BOILING AND PRESSUFILED WATER REACTOhS The NSSS vendor's practice and recommendations are in agreement with Regulatory Guide 1. 25, except that f ootnote C.1.c cannot ce met. This footnote states that the average turnup for the peak assemoly should be 25,000 Mwd / ton or less. The average ournup for the peak assemoly is in the 35,000 Mwd / ton range for Westinghouse fuel. (See Subsections 15.7.4. 2 and 15< 7. 4. 3. )

i A1.25-1 A;.W3.4407

i B/B-FSAR AMENDMENT 37 ,

MARCH 1982 REGULATORY GUIDE 1.26 l Revision 3, February 1976 QUALITY GROUP CLASSIFICATIONS AND STANDARDS FOR WATER , STEAM , AND RADIOACTIVE-CONTAINING COMPONENTS OF NUCLEAR POWER PLANTS The Applicant complies with the regulatory positions stated in this Regulatory Guide. Refer to Subsection 3.2.2 for further information.

A1.26-1 l

A0!'34408

B/B-FSAR AMENDMENT 43 SEPTEMBER 1983 QGULATORY GUIDE 1.27 Revision 2, January 1976 ULTIMATE HEAT SINK FOR NUCLEAR POWER PLANTS The Applicant meets all objectives set forth in this Regulatory Guide as presented in Subsections 2. 3.1, 2. 4.11, 9. 2. 5.1, 9. 2. 5. 2, l and 9.2.5.3.

J A1.27-1

{ A'3P2.%

B/B-FSAR AMENDMENT 37 MARCH 19 82 l

REGULATORY GUIDE 1.28 Revision 1, March 1978 QUALITY ASSURANCE PROGRAM REQUIREMENTS

{ (DESIGN AND CONSTRUCTION)

The Applicant complies with this regulatory guide. Refer to Chapter 17.0 of the FSAR for further information on the Commonwealth Edison Quality Assurance Program.

i 4

1 l

l

,_ A7'14410

S/B-FSAR AMENDMENT 37 MARCH 1982 REGULATORY GUIDE 1.29 Revision 3, September 1978 SEISMIC DESIGN CLASSIFICATION The Applicant complies with the regulatory positions stated in this Regulatory Guide. Refer to Subsections 3.9.2.7 and 3.10.1.2.1 for further information.

l A1.29-1 A003.4411

B/B-FSAR AMENDMENT 37 MARCH 1982 REGULATORY GUIDE 1.30 Revision 0, August 11, 1972 QUALITY ASSURANCE REQUIREMENTS FOR THE INSTALLATION, INSPECTION, AND TESTING OF INSTRUMENTATION AND ELECTRIC EQUIPMENT The Applicant complies with this regulatory guide.

Refer to Subsections 3.11.2, 7.1.2.8, and 8.1.5 of the FSAR for further information.

A0014412 A1.30-1

B/B-FSAR AMENDMENT 45

JUNE 1984 REGULATORY GUIDE 1.31 ,

l i

Revision 1 7 June 19 73 i CONTROL OF FERRITE CONTENT IN STAINLESS STEEL WELD METAL 4

I The Applicant complies with the Regulatory Position with the following comments and exceptions keyed to paragraph numbers in the Position.

, 1.a The Applicant essentially complies, the minor difference being that instead of 125 maximum delta j ferrite for wrought structures, the Applicant has j required the ferrite content to be 5 - 15% for both

! duplex cast and wrought structures. The 155 maximum ferrite content restriction exceeds the 1974 edition i of ASME Section III, which has no upper limit on ferrite content.

I j 1.b The Applicant essentially complies, except that alternatively the chemical analysis may be obtained directly from the wire, consumable insert, etc.

Undiluted weld deposits may be analyzed when alloy contributions are obtained from a flux such as in covered electrodes, flux-cored wires, etc.

The Applicant complies with the 1974 edition of ASME Section III on sampling for chemical analysis. ,

i 1.d The Applicant does not comply with this requirement and instead requires the maximum interpass temperature not to exceed 350' F.  ;

i l 1.e Applicant does not comply. ,

l The Applicant has been specifying ferrite control for j stainless steel welding of nuclear power plant

components since 1969. This experience has confirmed research findings that when 55 minimum ferrite is

present, as now specified in Section III, Fissuring i is virtually eliminated.  ;

i j 4. The Applicant does not comply and maintains that to

require welds to be made from a single heat is j i

! excessively restrictive. It would not only limit I each weld to the use of one welding process, but also  :

1 \

A1.31-1 A0024413 j

B/B-FSAR AMENDMENT 18 3: - @-, JANUARY 1979

( to a single diameter filler wire or electrode. Gas tungsten arc welding using consumable insert would be impossible. Instead, the Applicant requires that each fabricator's QA Program maintain control that "all austenitic stainless steel filler metal will have 5 to,15% ferrite. . - .--- -

5. The Applicant and Westinghouse do not believe that it

.is necessary to measure the ferrite content in

, production welds provided that adequate quality

~ control is exercised through control of process

,2 variables as defined in paragraph C1 of the Regulatory Guide and as discussed above. .Tmple

~ indication of the delta ferrite level and quality of i

welds will be determined by testing the qualification

welds. The large number of welders evaluated and the

many qualification welds each must-make will provide rufficient data to ascertain the effectiveness and 5 quality of welds made on austenitic stainless steel.

= =-- .. . , _ _ .. ...: :.

6. The ., Applicant does not comply.

-3__

7 She Applicant does n'o't comply and maintains that the requirement of 5 to -15% ferrite determined by p

s magnetic measurement on a weld deposited is sufficient justification.to eliminate the usefulness o_f ~ p.roduction weld. monitoring..

~~ ~~

The position concerning the control of delta ferrite in the stainless steel welding is discussed in Subsections 3.8.1.2, I;8,.3.2.2, 3.8.4.2, 5.2.3, 5.3.1.4cc.and.6.1 ,1.1.

g,. ,,,

r l

i

, A1.31-2 AOO1M14 l .31-2

B/B-FSAR AMENDMENT 37 MARCH 1982 REGULATORY GUIDE 1.32 i

Revision 2, February 1977 CRITERI A FOR SAFETY-RELATED ELECTRIC POWER SYSTEMS FOR NUCLEAR POWER PLANTS The Applicant complies with the regulatory positions of this guide with the following exceptions / clarifications:

i Regulatory Position C.l.a.

See Applicant's Position on Regulatory Guide 1.93.

Regulatory Position C l.d.

See Applicant's Position on Regulatory Guides 1.6 and 1.75.

Regulatory Position C.l.e.

See Applicant's Position on Regulatory Guide 1.75.

Regulatory Position C.l.f.

See Applicant's Position on Regulatory Guide 1.9.

Regulatory Position C.2.a.

See Applicant's Position on Regulatory Guide 1.81.

i Regulatory Position C.2.b.

See Applicant's Position on Regulatory Guide 1.93.

4 i

i l

A1.32-1 A')U14415

1 B/B-PSAR AMENDMENT 37 MARCII 1982 i

i

[ REGULATORY GUIDE 1.33 a

Revision 2, February 1978 1,

, QUALITY ASSURANCE PROGRAM REQUIREMENTS I

(OPERATION)

The Applicant complies with this regulatory guide. Refer to Chapter 17.0 of the FSAR for further information on the 1

Commonwealth Edison Quality Assurance Program.

i ,

I i

4 i

)

i i

i i

I l

l 1

i i ,

l A0014416 A1.33-1 l-

- -. - - .- -. . - . - - . - - - . - - - . ..= . - -. . - .- . --.- _ - . _._ , -

B/B-E S AR AMENDMENT 37 MARCH 1982 REGULATORY GUIDE 1.34 Revision 0, December 28, 1972 CONI'ROL OF ELeCTFOSLAG AELD PROPERTIES The Apelicant complies with the regulatory position whenever the electroslag welding process is used for components made of ferritic or austenitic materials, however, electroslag welding is not used for equipment purchased on the Applicant's syecifications for Byron /Braidwood. (See Subsections 5.3.1.4 and 5.4.2.1.1 ror f urther information.)

l i

I I

i i

1 A0014417 A1.34-1

B/B-FSAR AMENDMENT 38 MAY 1982 REGULATORY GUIDE 1.35 .

Proposed Revision 3, April 1979 INSERVICE INSPECTION OF UNGROUTED TENDONS IN PRESTRESSED i CONCRETE CONTAINMENT STRUCTURES I

The plant design conforms to the regulatory position described in proposed Revision 3 (April 1979) of this regulatory guide.

Refer to Subsections 3.8.1.7.3.2 and 16.4.6.1.7.1.

4 l

W I

A1.35-1

<'I18 l - _ _ - . _ _ _ _ -__ . . _. . _ - _ . _ . _ . _ _ -. -

B/B-FSAR AMENDMENT 38 MAY 1982 ,

REGULATORY GUIDE 1.35.1 Proposed Revision 0, April 1979 DETERMINING PRESTRESSING FORCES FOR INSPECTION OF PRESTRESSED CONCRETE CONTAINMENT STRUCTURES The plant design conforms to the regulatory position described in proposed Regulatory Guide 1.35.1. Refer to FSAR Subsections

! 3.8.1.7.3.2 and 16.4.6.1.7.1.

1 1

i i

i AUU14419 A1.35.1-1 i

i l - - , , - --- - - . ,- - - , , - . , - - , - , , - - , - - - , , , - , - , - - - - -

o B/B-FSAR AMENDMENT 37 MARCH 1982 REGULATORY GUIDE 1.36 Revision 0, February 23, 1973 NONMETALLIC THERMAL INSULATION FOR AUSTENITIC STAINLESS STEEL The Applicant complies with this guide.

The NSSS vendor practice meets recpmmendations of Regulatory Guide 1.36 but is more stringent in several respects as discussed below. (Also see paragraph 4.5.2.4 for further information.)

The nonmetallic thermal insulation used on the reactor coolant

pressure boundary is specified to be made of compounded materials which yield low leachable chloride and/or fluoride concentrations. The compounded materials in the form of blocks, boards, cloths, tapes, adhesives, cements, etc., are silicated to provide protection of austenitic stainless steels against stress corrosion which may result from accidental wetting of the insulation by spillage, minor leakage, or other contamination from the environmental atmosphere. Each lot of insulation material is qualified and analyzed to assure that all of the

! materials provide a compatible combination for the reactor coolant pressure boundary.

The tests for qualification specified by the guide (ASTM C692-71 or RDT M12-IT) allow use of the tested insulation materials if no more than one of the metallic test samples cracks. Westinghouse rejects the tested insulation material if any of the test samples cracks.

The vendor procedure is more specific than the procedures suggested by the guide, in that the Westinghouse specification requires determination of leachable chloride and fluoride ions from a sample of the insulating materials. The procedures in the guide (ASTM D512 and ASTM D1179) do not differentiate between leachable and unleachable halogen ions.

In addition vendor experience indicates that only one of the three methods allowed under ASTM D512 and ASTM D1179 for chloride and fluoride analysis is sufficiently accurate for reactor applications. This is the " referee" method, which is used by Westinghouse. These requirements are defined in Westinghouse l Process Specification PS-83336KA.

1 i

AOU3.44;o A1.36-1

B/B-FSAR AMENTMENT 37 MARCH 19 82 t

REGULATORY GUIDE 1.37 Revision 0, March 16, 1973 i

QUALITY ASSURANCE REQUIREMENTS FOR CLEANING OF FLUID SYSTEMS AND ASSOCIATED COMPONENTS OF WATER-COOLED. NUCLEAR POWER PLANTS i

The Applicant complies with this regulatory guide. Refer to Subsections 5.2.3, 5.4.2.1.1, and 6.1.1.1 of the FSAR for further information.

i 4

f f

.s . .99 A0014421 l

A1.37-1

i .

i B/B-FSAR AMENDMENT 37

, MARCH 1982 REGULATORY GUIDE 1.38 Revision 2, May 1977 QUALITY ASSURANCE REQUIREMENTS FOR PACKAGING, SHIPPING, RECEIVING, STORAGE, AND HANDLING OF ITEMS FOR WATER-COOLED NUCLEAR POWER PLANTS The Applicant complies with this regulatory guide. Packaging, shipping, receiving, storage, and handling of PWR power plant equipment are covered by quality specifications based on ANSI N45.2.2-1972. These practices are audited for com-pliance in accordance with the Commonwealth Edison Quality Assurance Program. Refer to Chapter 17.0 of the FSAR for i further information on the Quality Assurance Program.

i 1

i

)

l

\

i l

A0014422 A1.38-1

B/B-FSAR AMENDMENT 37 MARCH 1982 REGULATORY GUIDE 1.39 Revision 2, September 1977 HOUSEKEEPING REQUIREMENTS FOR WATER-COOLED NUCLEAR POWER PLANTS The Applicant complies with this regulatory guide. Control of facility cleanness, care of material and equipment, fire prevention and protection, disposal of debris, protection of material and control of access are all standard practices of the station construction department during construction.

After transfer of the plant to the operating staff, these functions are the responsibility of the production department.

Normal management attention and periodic audits under the Commonwealth Edison Quality Assurance Program will provide the desired result at B/B. Independent audits by NRC Region III personnel also contribute to the effectiveness of good housekeeping practices. Refer to Chapter 17.0 of the FSAR for further information on the Quality Assurance Program.

i r

i A1.39-1 AWMU

l l

l I

B/B-FSAR AMENDMENT 37 '

MARCH 1982 REGULATORY GUIDE 1.40 Revision 0, March 16, 1973 QUALIFICATION TESTS OF CONTINUOUS-DUTY MOTORS INSTALLED INSIDE THE CONTAINMENT OF WATER-COOLED NUCLEAR POWER PLANTS NSSS Scope It is the Westinghouse position that motors inside containment comply with the qualification control requirements of Criterion i

III to Appendix B to 10 CFR 50. These requirements are satisfied by qualification as described in WCAP-8587 and its supplement which contains appropriate EQDP's (Equipment Qualification Data Packages) for Westinghouse supplied continuous duty motors within the containment. The Applicant is in compliance with the objectives of Regulatory Guide 1.40.

Non-NSSS Scope The Applicant complies with the requirements of Regulatory Guide 1.40 with the clarification to the Regulatory Position identified and justified below:

Regulatory Position Cl To the extent practicable, auxiliary equipment that will be part of the installed motor assembly should also be qualified in accordance with IEEE 334-1971.

Applicant's Position Comply with regulatory position, in that to the extent practicable, auxiliary equipment essential to the safety function of the installed motor assembly will be qualified in accordance with IEEE 334-1971.

Justification of Applicant's Position Nonessential auxiliaries have no safety function and should be excluded from the requirements.

A1.40-1 A0!dMA'21 1

.- .. - - - - - -- _ . ~ - _ . . . - _ - . - . - . -.

B/B-FSAR AMENDMENT 37

' MARCH 1982 REGULATORY GUIDE 1.41 4

Revision 0, March 1973 i

PREOPERATIONAL TESTING OF REDUNDANT ONSITE ELECTRIC POWER SYSTEMS TO VERIFY PROPER LOAD GROUP ASSIGNMENTS 1

The applicant complies with this Regulatory Guide. Refer to FSAR Subsection 8.1.8 for further information.

A1.41-1 003A425

B/B-FSAR AMENDMENT 37 MARCH 1982 REGULATORY OUIDE 1.4 2 No Current Issue INTth1M LIC6NSING POLICY ON AS LOW A5 PRACTICABLE FOR GASEOUE RADIOIODINd RZLEASES FROM LIGHT WATER-COOLED NUCLEAR Powr R F EAC10E3 This guide has been withdrawn.

't i

e 9

A1.42-1 I

B/B-FSAR AMENDMENT 37 MARCH 1982 REGULATORY GUIDE 1.43 Revision 0, May 1973 CON TROL OF STAINLESS STEEL WELD CLALOING OF LOW-ALLOY STEEL COMPONENTS The applicant complies with the requirements of this guide. Refer to Subsection 5. 3.1.4 for further information.

destinghouse meets the intent or Regulatory Guide 1.4 3 ty requiring qualification of any high neat input process, such as the suomerged-arc wide-strip welding process and the submerged-arc-6-wire process used on SA-508 Class 2 material, with a performar.ce test as described in hegulatory Position 2 of the guide. No qualifications are required by the regulatory guioe for SA-533 material and equivalent chemistry f or forging grade SA-506 class 3 material.

AOUSA427 A1.43-1 i

l l

B/B-FSAR AMENDMENT 40 NOVEMBER 1982 ,

l REGULATORY GUIDE 1.44 Revision 0, May 1973 i

CONTROL OF THE USE OF SENSITIZED STAINLESS STEEL l

The Applicant complies with this guide with the following j clarifications keyed to paragraph numbers in the Regulatory Position.

r 1. The Applicant complies in general with the intent of l the requirements of this guide. With regard to fabrication, shipment, storage, and construction, the Applicant requests that contaminants be avoided and cleaning solutions be halide free.

2. The Applicant complies with the requirement in that l '.

it specified ASME material specifications which require material to be supplied in the solution annealed condition.

3. The Applicant does not agree with this requirement.  !

Specification of solution annealed material is sufficient.

4. The Applicant's specifications prohibit the use of l materials that have been exposed to sensitizing temperatures in the range of 800' to 1500' F.

5. Same as Item 4.

g

6. The Applicant will incorporate the requirement that l all welding procedures for 3XX Series of materials include an intergranular corrosion test to be

, submitted for each welding procedure.

, The position on Regulatory Guide 1.44 is discussed in part in subsection 5.2.3.4 (Fabrication and Processing of Austenitic Stainless Steel) and in Subsection 6.1.1.1.

1 l

4 l

gj,gg.j hb.

e-ei--y- rye +- .e ,.y -w

B/B-FJAR AMENDMENT 37 tiARCH 19 82 REGULATORY GUIDE 1.45 Revision 0, May 1973 REACIOR COOLANI PRESSURE BOUNDARY LtAKAGE DETECTION SiSTEMS The Applicant complies with this guide with the following I clarifications keyed to paragraph numbers in the Regulatory Position.

1. Identified leak sources will be piped to either the EC drain tank or a miscellaneous drain tank to ce utilized for this purpose only.

Temperature of selected drain lines will be monitored to identify leaks. Tank inventories will te monitored. Iemperature monitoring is more sensitive to small leaks than tiow rate monitoring specified in the PoEition.

2. Unidentified leak sources will ce monitored to as accurate an equivalent flow rate as is practicable.

No guarantee can De made, at this time tnat a 1 gpm accuracy can be maintained for all unidentified leakage monitors.

3. The following leax detection systems will os provided:

Identified Sources

1. RC drain tank level indication and temperature indication of selected inlet lines, or

2. miscellaneous drain tank level indication and temperature indication of selected inlet lines.

Unidentified Sources

1. containment sump level, T

2. airborne particulate activity monitoring, and A1.45-1 AW34423

B/B-FSAR

3. airborne gaseous radioactivity monitorir.g.

l

4. Intersystem leakage between primary and secondary  ;

plant will be monitored via air ejector off gas '

radiation monitors. Also, pressurizer and makeup tank levels will be monitored to yield total reactor coolant leakage.

5. Leak detector sensitivity will be as low as practicable.

6. Conversions to common leakage equivalent will be supplied to operators wherever possible. Conversions to a common leakage equivalent is not possible in all cases. In these cases, the system is intended primarily for localization or identification of a leak with no quantitative implications.

Further information on reactor coolant pressure boundary leak detection can be found in subsection

5. 2. 5.

A1.45-2 Ane3M30

B/B-FSAR AMENDMENT 37 MARCH 19 82 i REGULATORY GUIDE 1.46 Revision 0, May 1973 PROTECTION AGAINST PIPE WHIP INSIDE CONTAINMENT i The applicant complies with Regulatory Guide 1.46. Further clarifications are provided in Subsections 3.6.2.3, 3.9.3.4.2, 3.8.3.2.1, 5.4.11.3, and 7.1.2.10.

i l

I l A1.46-1 A0014431

B/B-FSAR A!!ENDMEriT 37 MARCH 1982 i

REGULATORY GUIDE 1.47 Revision 0, May 1973 BYPASSED AND INOPERABLE STATUS INDICATION FOR NUCLEAR POWER PLANT SAFETY SYSTEMS The Applicant complies with the Regulatory Position of this guide as

discussed in FSAR Section 7.5 and Subsections 8.1.9 and 7.1.2.10.

4 l

l A003.4432 A1.47-1

B/B-FSAR AMEND!!ENT 37 MARCH 1982 i

REGULATORY GUIDE 1.48

\

Re sion 0, May 1973 i "O I

DESIGN LIMITS AND LOADING COMBINATIONS FOR SEISMIC CATEGORY I FLUID SYsTsd COMPONENTS 4

The Applicant complies with this guide with the following l

[

clarifications keyed to paragraph numbers of the guide: -

A a. Regulatory Position 4.a-- The stress limits of NB-3222 for the normal, upset, and emergency conditions, and the stress limits of NB-3223 for the s faulted condition, are more realistic criteria.

b. Regulatory Position 8.a - For the emergency condition, a stress limit of 1.8Sh is more realistic.

j c. Regulatory Position 8.b - For the faulted condition, a stress limit of 2.4 S h is more realistic.

d. Regulatory Position 9.a - For emergency conditions, the following stress criteria are more realistic:

. membrane stress, 1.5 S; membrane stress plus bending stress, 1.8 S.

e. Regulatory Position 9.b - For loadings associated i with the faulted conditions plus normal station I conditions plus SSE, the following stress criteria-l are more realistic: membrane stress, 2.0 S; membrane stress plus bending stress, 2.4 S.

f. Regulatory Position 10.a - For loadings associated j with normal conditions plus SSE dynamic system i loadings associated with faulted conditions, the i following stress criteria are more realistic:

membrane stress, 1 1.2 S; membrane stress plus

bending stress, 1 1.8 S.

g. Regulatory Position - 11.a - For the normal condition or the upset condition, it is more realistic that the primary-pressure rating Pr should not be exceeded by more than 10%. For the emergency condition, it is more realistic that Pr should not be exceeded by more than 20%.

A1.48-1 A003.4433

l B/B-FSAR AMENDMENT 20 ,

MAY 1979 l

h. Regulatory Position 11.b - For the loadings associated with the normal condition, the vibratory motion of the SSE, and the dynamic system loadings associated with the faulted conditions, it is more realistic that Pr should not be exceeded by more than l 50%.

i. Regulatory Position 12.a - For loadings associated with normal conditions plus SSE plus dynamic system loadinas associated with faulted conditions, it is more rialistic that Pr should not be exceeded by more l than 105.

Further discussion on this subject can be found in Sections 3.7, 3.8, and 5.2.

i A0014431 A1.48-2

. s b/B-FSAR AMENDMENT 37 MARCH 19 82 REGULATCRY GUIDE 1.49 Revision 1, December 1973 POWER LEVELS OF NUCLEAR POWER PLANTS Tne Byron /braidwood design meets the recommendations of Regulatory Guide 1.49, since tne projected initial power level is less than 3800 Et and analyses and evaluation are made at assumed core power levels less than the levels in the guide.

i l

l i

A1.49-1 A;)D3.4435 l

l

B/B-FS AR AMENDMENT 37 MARCH 1982 REGULATORY GUIDE 1.50 Revision 0, May 1973 CONTROL CF PREHEAT TEMPERATURE FOF WELDING OF LOW-ALLOY STEEL The Applicant complies with this guide with the following comments and exceptions keyed to paragraph numbers in the Regulatory Position.

4 1.a. Applicant requires preheat temperatures referenced in applicaole codes but does not require a maximum interpass temperature. To date, it has not oeen founo necessary to specity a maximum interpass temperature.

1.b. Welding procedures will ne qualified in tne preheat temperature range. It is not possicle to consistently maintain preheat at the minimum temperature during welding procedure qualification.

2. The Applicant does not agree with this positior..

It is an impossibility to maintain preheat tem eerature during f abrication of spool pieces when tour or five welds have been made prior to a complete post-weld heat treatment of the spool piece. The only way this could be accomplished would be to have intermittent post-weld heat treatment which in the case of the higher alloy steel, such as 2-1/4 chrome, may be detrimental.

3. Preheat temperature limit will be monitored, cut not interpass temperature.

Westinghouse considers that this Guide applies to ASML dection III class 1 Components.

The NSSS vendora's practice for Class 1 components is in

agreement with the requirements or Regulatory Guide 1.50 except for regulatory positions 1 (b) and 2. for class 2 anc 3 components, Westinghouse does not apply any of Regulatory Guice 1.50 recommendations.

l A1.5o-1 A3034436

B/B-FSAR In the case of regulatory position 1 (b), the welding procedures are qualified within the preheat temperature ranges required oy Section IX of the ASME Code. Westinghouse qualification procedures.

In the case of regulatory Position 2, the vendor's position described in WCAP-8577, "The Application of Preheat Temperature After Welding of Pressure Vessel Steels," has been found acceptable oy the NRC. This WCAP establishes the guidelines which permit the component manufacturer to either maintain the preheat until a post-weld heat treatment or allow the preheat to drop to ambient temperature.

In the case of reactor vessel main structural welds, the practice of maintaining preheat until the intermediate or final post-weld heat treatment has oeen followed by the vendor. In either case, the welds have shoen high integrity The NSSS vendor meets Regulatory Position 4 in that, for their components, the examination procedures required by Section III and che inservice inspection requirements of Section XI are met.

(For further information, see Paragraph 5.3.1. 4.)

l l

l l

I I

(

ADt' M G 7 M.9-2

B/ B-FS AE AMENDMENT 37 MARCH 19 82 REGULATORY GUIDE 1.51 No Current Issue H SERVICE INSPECTION OF ASME CODE CLASS 2 AND 3 NUCLEAR POWER PLANT COMPONENTS This guide has oeen withdrawn.

l 1

1 i

i 1

A1. 51 -1 AOU14438

i . .

l B/B-FSAR AMENDMENT 45 JUNE 1984 REGULATORY GUIDE 1. 52 Revision 2, March 1978 DESIGN, TESTING AND MAINTENANCE CRITERIA FOR ENGINEERING-SAFETY-FEATURE ATMOSPHERE CLEANUP SYSTEM AIR FILTRATION AND ADSORPTION UNITS OF LIGHT-WATER-COOLED NUCLEAR POWER PLANTS

~

The Applicant complies with the Regulatory Position with the following comments and exceptions keyed to paragraph numbers i in Section C of the Position:

1.e (Deleted) (Note 1) 2.a Entrained water droplets are not considered credible due to significant quantities of ductwork with elbows.

, Water droplets, if present, will impinge on ducts and drop out of vertical duct risers as the air enters exhaust plenums. However, the auxiliary building exhaust system does contain prefilters which can serve as demisters.

2.d (Deleted) (Note 1) 2.f The auxiliary building non-accessible area exhaust filter system consists of three built-up filter trains (one standby) with a capacity of 66,900 cfm each. The efficiency of the filters will be tested at this capacity. For maintenance purposes each train is divided into three banks with a concrete floor between each bank. Each filter bank is three filters high and seven wide.

The auxiliary building accessible area exhaust filter system has a capacity of 162,300 cfm. The efficiency of the filters will be tested at this capacity. For maintenance purposes the system is divided into twelve built-up filter banks (three standby). Each filter bank is three filters high and seven wide. This filter system is designed in accordance with Regulatory Guide 1.52; however, it will be in-place tested in accordance with the requirements of Regulatory Guide 1.140.

29 All ESF filter systems have local control panel airflow indication. In addition, the flow rate of the main auxiliary building exhaust' fans is recorded at the local control panel. Airflow through the control room emergency makeup air filter units and the auxiliary building charcoal booster fans is continuously sensed and flow rate is controlled to maintain constant airflow.

A1.52-1 i

O O

. 1 B/B-FSAR AMENDMENT 42 MAY 1983 The differential pressure across all of the ESF filter unit fans is indicated on local control panels. High and low differential pressure, and fan trip annunciation is provided on the main control panel.

The differential pressures across all HEPA filters are indicated on local control panels. In addition, the ,

pressure drop across HEPA t'lters upstream of charcoal filters is recorded in the acin control room. High differ-ential pressure across all HE.'A filters is annunciated on the main control panel and on local control panels.

2.j Filter trains are not designed to be removable f rom the building as an intact unit. Tie size of the train precludes shipment off-site and there ari no facilities for on-site disposal of the intact unLt. Thc filter elements are removable and can be dispcied of through the solid radwaste system.

2.1 Filter Housings All of the auxiliary building and fuel handling building exhaust system filter housingc are designed in accordance with ANSI N509-76. The housings are at negative pressure

, with respect to their surroundings and are located in auxiliary building general area which is a low airborne radiation environment. Any in-leakage from the general area will not adversely affect Appendix I releases.

Hence, the housings will not be leak tested to the ANSI N509-76 requirements.

The control room emergency makeup air system filter housings are designed in accordance with ANSI N509-76. The filter housings are at negative pressure with respect to their surroundings, and are located within the control room boundary which is a habitable environment the same as the control room. Any in-leakage will be from the control room environment and, therefore, will not adversely affect the quality of that environment; hence, the housings will not be leak tested to ANSI N509-76 requirements.

Ductwork All auxiliary building and fuel handling buildings exhaust system ductwork upstream of the filter units is under negative pressure with respect to its surroundings and is located in the same areas of the buildings served by the exhaust systems. Any in-leakage will be filtered prior to discharge to the atmosphere, hence, this ductwork will not be tested to ANSI N509-76 requirements.

l A1.52-2 A0014440 t

B/B-FSAR AMENDMENT 42 MAY 1983 All control room emergency makeup air system ductwork is located within the control room boundary which is a habitable environment. Any ductwork leakage will not adversely affect the habitability of the environment, hence, this ductwork will not be tested to ANSI N509-76 requirements.

The design airflow quantities for each system will be verified during testing, adjusting and balancing of the systems. Deviations of more than + 10% of the design flow quantities will be evaluated and any disposition will be documented.

2.m (Deleted) (Note 1) 3.b (Deleted) (Note 1) 3.d (Deleted) (Note 1) 3.e (Deleted) (Note 1) 3.h (Deleted) (Note 1) 3.1 (Deleted) (Note 1) 4.b The space provided between components is 3 feet from the front (or rear) of the components to the nearest obstacle (filter frame or other filter component). This allows 3 feet of access between components.

4.c (Deleted) (Note 1) 5.b Airflow distribution tests will be performed to ensure that the airflow through any individual filter element does not exceed 120% of the elements rated capacity.

5.c (Deleted) (Note 1)

Further discussions on this subject can be found in Subsections 9.4.1.2 and 12.3.1.7.

Note 1: Exception to this section is no longer required because the Regulatory Guide has been revised to eliminate the criteria to which exception was originally taken.

A1.52-3 9g441

B/B-FSAR A!!END11ENT 37 MARCH 19 82 REGULATORY GUIDE 1.53 Revision 0, June 1973 APPLICATION OF THE SINGLE-FAILURE CRITERION TO NUCLEAR POWER PLANT PROTECTION SYSTEMS The Applicant complies with the guidelines tor application of the single f ailure criteria to nuclear power plant protection systems as discussed in Subsections 7.1.2.11 and o.1.10.

l A1. 53- 1 .

g442

B/B-FSAR AMENDMENT 37 MARCH 1982 REGULATORY GUIDE 1. 54 Revision 0, June 1973 QUALITY ASSURANCE REQUIREMENTS FOR PROTECTIVE COATINGS APPLIED TO WATER-COOLED NUCLEAR POWER PLANTS i Ihe Applicant complies with this Regulatory Guide Position. (see Suosection 6.1.2 for furtner inf ormation.)

The NSSS equipment located in tne containment Duilding is se e arated into four categories to identify the applicability of this regulatory guide to various types of equipment. These categories of equipment are as follows:

Category 1 - Large equipment Category 2 - Intermediate equipment Category 3 - Small equipment Category 4 - Insulated / stainless steel equipment The equipment in each of these categories is identified below.

Category 1 - Large Equipment Since Category 1 equipment occupies a large surf ace area (17,230 square feet, 4.1% of containment surf ace area) anc is procured from only a f ew vendors, stringent requirements for protective coatings are specified through painting process specifications in procurement documents. These process specifications define requirements for:

a. Preparation of vendor procedures.

o. Use of specific coating systems which are qualified to ANSI N101.2 " Quality Assurance for Protective coatings Applied to Nuclear lacilities".

c. Surf ace preparation.

d. Application of the coating systems in accordance with the paint manuracturer's instructions.

l A1.54-1 f003,4443

B/B-FSAR AMENDMENT 21 JULY 1979

e. Inspections and non-destructive examinations.

. f. Exclusion of certain materials,

g. Identification of all nonconformances.

h. Certifications of compliance.

The vendor's procedures are subject to review by PWRSD Engineering personnel, and the vendor's implementation of the specification requirements is monitored during the Westinghouse QA Surveillance activities.  :

. This system of controls provides assurance that the protective coatings will properly adhere to the base metal during prolonged exposure to a post-accident environment present within the containment building.

Category 2 - Intermediate Equipment The total exposed surface area of Category 2 equipment is approximately 3450 sq. ft. (.8% of containment surface area).

Since these items are procured from a large number of vendors and individually occupy very small surface areas, it is not practical to enforce the complete set of stringent requirements which were applied to Category 1 items. However, Westinghouse does implement another process specification in procurement documents.

This specification define,s to the vendors the requirements for:

a. Use of specific coating systems which are qualified to ANSI N101.2, " Quality Assurance for Protective l

. Coatings Applied to Nuclear Facilities."

b. Surface preparation.

c. Application of the coating systems in accordance with the paint manufacturer's instructions.

The vendor's compliance with the requirements is also checked during the Westinghouse QA Surveillance activities in the vendor's plant. These measures of~ control provide a high degree of assurance that the protective coatings will adhere properly to the base metal and withstand the postulated accident environment within the containment building.

, Category 3 - Small Equipment I

Category 3 items are procured from several different vendors and are painted by the vendor in accordance with conventional industry practices. Because the total exposed surface area is only 900 sq. ft., Westinghouse does not specify further requirements.

A0014444 A1.54-2 I

l

D/B-FSAR AMENDMENT 18 JANUARY 1979 Category 4 - Insulated or Stainless Steel Equipment Since Category 4 equipment is insulated or is stainless steel, no painted surface areas are exposed within the containment.

Therefore, this regulatory guide is not applicable for Category 4 equipment.

Category 1 - Large Equipment

a. Reactor Coolant System Supports Category 2 - Intermediate Equipment

a. Seismic Platform and Tie Rods

b. Reactor Internals Lifting Rig

c. Head Lifting Rig

d. Electrical Cabinets Category 3 - Small Equipment

a. Transmitters

b. Alarm Horns

c. Small Instruments

d. Valves

e. Heat Exchanger Supports Category 4 - Insulated or Stainless Steel Equipment

a. Steam generators - covered with metallic reflective l insulation

b. Pressuriser - covered with metallic reflective l insulation

c. Reactor Pressure Vessel - covered with metallic l reflective insulation

d. Reactor Coolant Piping - stainless steel

e. Reactor Coolant Pump Casings - stainless steel l

l Aac14445 A1.54-3

B/B-FSAR AMENDMENT 38 MAY 1982 i

REGULATORY GUIDE 1.55 Revision 0, June 1973 J

CONCRETE PLACEMENT IN CATEGORY I STRUCTURES The plant design conforms to the regulatory position with the following exceptions:

1. ACI 301-72 specifies that the frequency for cylinder testing shall be two cylinders per 100 yards of concrete, tested at 28 days with a minimum of one set per day for each class of concrete.

The Applicant's position is to use six cylinders per 150 yards of concrete, tested at 7, 28, and 91 days, with a minimum of one set per day for each class of concrete. This exceeds the requirements of both ACI 318-77 and ACI 349-76.

The compliance with the requirements of this regulatory guide is detailed further in the response to Questions 130.21 and 421.19.

2. ACI 301-72, subsection 8.5.3, requires that grouting be applied on the vertical surfaces of construction joints. This requirement has been removed from ACI 381-80.

l l

I t

A1.55-1 A003.4446

B/B-PSAR AMENDMENT 37 MARCH 1982 REGULATORY GUIDE 1.56 Revision 1, July 1978 MAINTENANCE OF WATER PURITY IN BOILING WATER REACTORS This Regulatory Guide is pertinent to BWR's only.

l l

l 1

A1.56-1 l

A003.tl.77

B/B-FS AR AMENDMENT 37 tiARCH 19 82 REGULATORY GUIDE 1.57 j Revision 0, June 1973 I

DESIGN LIMI1S AND LOADING COMBINATIONS FOR MEZAL_PRIMARr REACIOR CONI AINMENT SYSTEM COMPONEN1S Commonwealth Edison complies with the hegulatory Position with the following clarifications.

Piping penetration assemolies will ce designed by the following guidelines:

a. The portion of tne primary containment penetration assembly which is part of the containment coundary, i.e., the penetration sleeve in its entire length (including the sleeve projection that forms an extension to the wall) , will be designed in accordance with Subsection NE,Section III of the ASME Code, augmented by the applicacle provisions of Regulatory Guide 1.57.

c. Tne portion of the primary containment penetration assemoly which consists of the heaa fitting (flued nead) and part of the process pipeline, will be designed in accordance with Subsection NB of the Code so as to satisfy stress requirements for design conditions (NB-3112, NB-2 221) , normal and upset l

conditions (NB- 3113.1, NB-3112.2, NB-32 22, NB-3223) ,

emergency conditions (NB-3224) , f aulted conditions l (NB-3113.4, F-1324.1, F-1324.6, Table F-1322) , and I

testing conditions (No-32 26, NB-6222, NB-6 322) .

Part b of the Applicant's e osition, whicn ref ers to the NB classification of the flued head and process pipe, is supported cy NA-2134 of the Code and note 3 of Regulatory Guide 1.57.

A00M448

B/B-FSAR AMENDMENT 33 OCTOBER 1981 i

REGULATORY GUIDE 1.58 Revision 1, September 1980 00ALIFICATION OF NUCLEAR POWER PLANT INSPECTION, EXAMINATION, AND TESTING PERSONNEL The applicant complies with the positions of this Regulatory Guide with the following exception:

Regulatory-Guide 1.58, Revision 1, Position 6 requires that a candidate for Level I, II, or III Inspector for inspection activities be a high school graduate or should have earned the General Education Development equivalence of a high school diploma. In the applicant's judgment, this is an unnecessary and unfair restriction. Personnel can be trained, evaluated, and tested to determine if they are qualified for a specific test or inspection function, regardless of whether or not they have a high school diploma or the equivalent. A person should be utilized based upon his capability to perform the job, and on his qualifications and abilities. In the applicant's view, continual on the job training with timely performance reviews is an acceptable method of complying with 10 CFR 50, regarding such qualification of plant inspection, examination, and testing personnel.

Also refer to the Commonwealth Edison Company Quality Assurance Program Topical Report CE-1-A.

l l

I l

A0014443 A1.58-1

AMENDMENT 28 B/B-FSAR OCTOBER 1980 d

(2)

High school graduation plus three years of relate i experience in equivalent inspection, examinat on, or testing activities, or (3)

Completion of college level work leading to an  !

Associate Degree in a related discipline plus one year related experience in equivalent inspection, examination, or testing activities, or (4)

Four-year college graduation plus six months of related experience in equivalent inspection, examination, or testing activities.

"3.5.3 Level III (1) Six years of satisfactory performance as a Le category or class, or d (2)

High school graduation plus ten years of relate experience in equivalent inspection, examination, or testing activities; or high school graduation l plus eight years experience in equivalent inspection, l examination, or testing activities, atwith at leastt associated with nuclear facilities - or if not,least relevant quality assurance aspects of a nuclear facility, or (3) Completion of college level work leading to anAss ence in equivalent inspection, examination, ortest experience if not, associated with nuclear facilities - orat le with the relevant quality assurance aspects of a nuclear facility, or (4)

Four-year college graduation plus five years of re-lated experience in equivalent inspection, examina-tion, or testing activities, with at least two years at least sufficient training to be ac-

- or if not, quainted with the relevant quality assurance aspects!I of a nuclear facility."

• During the construction phase, Commonwealth Edison uses ,

previous performance and satsifactory completion of capa- ific bility testing as an acceptable alternative to the spec years of education or experience.

AOPSA450 A1.58-2

B/B-FSAR AMENDMENT 32 AUGUST 1981 i

Specific exceptions are covered in the 422 question series responses. Offsite inspection, examination and testing personnel are qualified in accordance with ANSI N45.2.6-1973 with the exceptions noted above.

I Commonwealth Edison complies with the objectives set forth in the referenced revision of this regulatory guide, as indicated in the textual material mentioned above.

For work within Westinghouse scope performed for the Byron and Braidwood plants, before January 1, 1975, the qualifica-tion of inspection, examination, and testing personnel was controlled by standard industry practice with the require-mants of SNT-TC-1A. The qualification of inspection, exam-ination, and testing personnel was accomplished locally by Westinghouse suppliers through on-the-job training.

For activities initiated on the Byron and Braidwood plants i

after January 1, 1975, Westinghouse follows the guidance

of this Regulatory Guide as defined in WCAP-8370 " Westinghouse Nuclear Energy Systems Divisions Quality Assurance Plan,a Revision SA and as described below.

This guide recognizes ANSI N45.26-1973 " Qualifications of Inspection, Examination, and Testing Personnel for the Con-struction Phase of Nuclear Power Plants." Westinghouse

policies and procedures for qualification of personnel en-gaged in inspection, examination and testing activities follow the guidance of this standard. Westinghouse uses demonstrated capability of performing the assigned tasks to predetermine standards or levels or proficiency as the primary basis for evaluating and certifying the personnel as an ac-ceptable alternative to the specific years of education /

experience.

1 Westinghouse applies the guidance of this standard to per-I sonnel who perform inspection, examination and testing activ-ities including surveillance of these activities for safety-related equipment, materials and services in the NES Divisions and at outside suppliers.

A0014451 A1.58-3

B/B-FSAR AMENDMENT 37 )

MARCH 1982 REGULATORY GUIDE 1.59 4

Revision 2, August 1977 DESIGN BASIS FLOODS FOR NUCLEAR POWER PLANTS The plant design conforms to the regulatory positions as de-i scribed in Subsection 2.4.3.

i J

f A1,59-1 A3014452

B/B-FSAR AMENDMENT 37 MARCH 1982 REGULATORY GUIDE 1.60 l Revision 1, December 1973 DESIGN RESPONSE SPECTRA FOR SEISMIC DESIGN OF NUCLEAR POWER PLANTS The plant design conforms to the regulatory positions as de-scribed in Subsection 2.5.2.

j A1.60-1 A0024453 l l

B/B-FSAR AMENDMENT 37 MARCH 1982 REGULATORY GUIDE 1.61 Revision 0, October 1973 DAMPING VALUES FOR SEISMIC DESIGN OF NUCLEAR POWER PLANTS The plant design conforms to the regulatory positions as de-scribed in Subsections 3.7.1.3 and 3.7.3.14, with the single exception of the large piping systems (diameter greater than 12 inches) SSE condition value of 3% critical. A conservative value of 4% critical for the Westinghouse reactor coolant loop configuration has been justified by testing and has been approved by the NRC staff. The test results are given in WCAP-7921-AR,

" Damping Values of Nuclear Power Plant Components." The use of higher damping values, when justified by documented test data, have been provided for in Regulatory Position C2.

l

('

Al.61-1

~0014454 A

i

u B/B-FSAR A!!ENDIiENT 37 MARCH 1982

- REGULATORY GUIDE 1.62 Revision 0, October 1973 4

MANUAL INITIATION OF PROTECTIVE ACTIONS The Applicant complies with this Regulatory Guide. Refer to FSAR Section 7.3, Subsection 8.1.11, and Paragraphs 7.1.2.1.2 and 7.2.1.1.2 for further information.

4 l

l A0c3.4455 l

A1.62-1

l i

l B/B-FSAR AMENDMENT 46 JANUARY 1985 REGULATORY GUIDE 1.63 Revision O', October 1973 ELECTRIC PENETRATION ASSEMBLIES IN CONTAINMENT STRUCTURES FOR WATER-COOLED NUCLEAR POWER PLANTS Regulatory Guide 1.63 supplements IEEE 317-1972, IEEE Standard for" Electric Penetration Assemblies in Containment Structures for Nuclear Power Generating Stations", and contains no specific testing recommendations. Regulatory Position C-4 does however add the quality assurance requirements of ANSI N45.2-1971 and ANSI N45.2.4-1972 to Section 8 (Required Data and Quality Control and Quality Assurance Procedures) of IEEE 317-1972.

The design, construction, installation, and testing of the electrical penetration assemblies will be in accordance with the quality assurance requirements of Regulatory Position C-4. (See Subsections 3.11.2, 7.1.2.13, and 8.1.12 for further information.)

l l

1 l

A0014456

B/B-FSAR AMEND!!ENT 37 MARCH 1982 REGULATORY GUIDE 1.64 Revision 2, June 1976 QUALITY ASSURANCE REQUIREMENTS FOR THE DESIGN OF NUCLEAR POWER PLANTS Regulatory Guide 1.64 describes an acceptable method of complying with the NRC's quality assurance requirements for the design of nuclear power plants. The Applicant complies with this regulatory guide. Refer to Chapter 17.0 of the FSAR for further information on the Commonwealth Edison Quality Assurance Program.

i i

l A0011457 l

A1.64-1

B/B-FSAR AMENDMENT 37  ;

MARCH 1982 RLGULATORY GUIDE 1.65 l Revision 0, October 1973 MATZRIALS AND INSPECTIONS FOR REACTOR VESSEL CLOSURE STUDS The Apslicant complies with Regulatory Guide 1.65 except for material and tensile strength guidelines.

Westinghouse has specified both 45 It-lo and 25 mils lateral expansion for control of fracture tougnness determined by Charpy-V testing, requirea by the ASME Boiler anc Pressure Vessel Code, dection III, Summer 1973 Addenda and 10 CER 50, Appenaix G (JULY 1973, Paragraph IV.A.4). These toughness requirements assure optimization of the stud bolt material tempering operation with the accompanying reduction of the tensile strength level when compared with previous ASME Boiler and Pressure Vessel Code requirements.

Ins specification or coth impact and maximum tensile strength as stated in the guide results in unnecessary hardship in erocurement of material without any additional improvement in quali ty. The closure stud Dolting material is procured to a minimum yielo strength of 130,000 psi and a minimum tensile strength or 145,000 psi. Inis strength level is compatible with the fracture toughness requirements of 10 CFP 50, Appendix G (JULY 1973, Paragraph 1.C), although nigner strength level colting materials are permitted oy the code. Stress corrosion has not oeen ooserved in reactor vessel closure stud bolting manuracturea from material of this strength level. Accelerated stress corrosion test data do exist for materials of 170,000 psi minimum yield strength exposed to marine water environments stressed to 75% of the yield strength (given in Reference 2 of the guide) . Tnese data are not considered applicable to destinghouse reactor vessel closure stud bolting because of the specified yield strength differences and a less severe environment; this has been demonstrated oy years of satisf actory service experience.

The ASME aoiler and Pressure Vessel Code requirement for toughness for reactor vessel Dolting has precluded the guice's

! additional recommendation for tensile strength limitation, since t

to ontain the required toughness levels, the tensile strength levels are reduced. Prior to 1972, the Code required to 35 ft-lo toaghness level whicn providea maximum tensile strength levels ranging from approximately 155 to 178 kesi (Westinghouse review A1.65-1 A0014458

B/B-FSAR of limited data - 25 heats) . After puolication of the Summer 1973 Addenda to the Code and 10 CFh 50, Appendix G, wherein the toughness requirements were modified to 45 ft-lb with 24 mils lateral expansion, all colt material data reviewed on Westinghouse plants showed tensile strengths of less than 170 kpsi.

Additional protection against the possibility or incurring corrosion effects is assured by:

1. Decrease in level of tensile strength comparable with the requirement of fracture toughness as described above.

2. Design of the reactor vessel studs, nuts, and washers, allowing them to be completely removed during each refueling permitting visual and/or nondestructive inspection in parallel with refueling operations to assess protection against corrosion, as part of the inservice inspection program.

3. Design of the reactor vessel studs, nuts, and washers, providing protection against corrosion by allowing them to be completely removed during each refueling and placed in storage racks on the containment operating deck, as required by Westinghouse refueling procedures, The stud holes in the reactor flange are sealed with special plugs before removing the reactor closure. Thus, the bolting materials and stud

~

holes are never exposed to the oorated refueling cavity water.

4. Use of a manganese phosphate surface treatment.

Use of Code Case 1605 does not constitute an issue between the NRC and Westinghouse inasmuch as (a) no questions have Deen raised on this point in vendor's approved standard reference document discussions of this Guide and (b) use of this code case has been approved oy the NRC via the guideline of Regulatory l Guide 1.85 (see Revision 6, May 1976) .

Further discussion of Reactor Coolant Pressure Boundary Materials, Inspection, and Testing is in Subsections 5. 2. 3 and 5.2.4 of the FSAR.

i l

A1. 65-2

B/B-FS AR A'4ENDI1ENT 37 MARCH 1982 RcGULAiORY GUIDE 1.66 No Current Issue l

NONDESTRUCTIVE EXAMINATION OF TUBULAR PRODUCIS This Regulatory Guide has oeen withdrawn (September 28, 1977).

i l

a 1

t

,h i

i  !

AOP14460 l A1.66-1  ;

t

AMENDMENT 43 B/B-FSAR SEPTEMBER 1983 REGULATORY GUIDE 1.67 No Current Issue II:0TALLATION OF OVERPRESSURE PROTECTION DEVICES This Regulatory Guide has been withdrawn (April 15,1983) .

I A1.67-1 A903' FIG 1 f

a - - -- - - - - . . ~ - .

' B/B-FSAR AMENDMENT 46 JANUARv 1985 REGULATORY GUIDE 1.68 Revision 2, August 1978 i

INITIAL TEST PROGRAMS FOR WATER-COOLED REACTOR POWER PLANTS The Applicant complies with this regulatory guide, as described in Chapter 14.0, with the following exceptions:

Appendix A.4.t, states " Performance of natural circulation tests .

of the reactor coolant system to confirm that the design heat l removal capability exists or to verify that- flow (without pumps) or temperature data are compatable to prototype designs for which equivalent tests have been successfully completed (PWR) . "

As described in the Byron SER the Applicant will be referencing the natural circulation testir.g which will be performed at Diablo

! Canyon. A preliminary assessment of differences between Byron and Diablo Canyon that may affect boron mixing under natural circulation has been provided and indicates that the Diablo Canyon test results and supporting analysis would satisfy the necessary requirements for Byron. If the tests at Diablo Canyon are not completed or the test results are unsatisfactory, the Applicant

, will perform such tests prior to startup after the first refueling for Byron. Additionally, simulator training for Byron reactor operators includes natural circulation procedures training.

Appendix A.S.a states " Determine that power reactivity i coefficients (PWR) or power vs. flow characteristics (BWR) are in accordance with design values (25%, 504, 754, 1004)."

Per recommendations of Westinghouse, Byron NSSS vendor, the Applicant intends to perform this testing at the 304, 504, 754, and 904 power ascension testing plateaus. These testing plateaus correspond to those previously listed in Table 14.2-82 for the Power Reactivity Coefficient Measurement Startup Test.

Appendix A.5.h states " Check rod scram times from data recorded during scrams that occur during the startup test phase to deter-mine that the scram times remain within allowable limits."

During power ascension testing, the Applicant does not intend to formally instrument the rod position indication system for reactor trip review purposes because this would require removal of the rod position indication from service. This would be a violation of Technical Specifications. For this reason the Appli-cant believes that this requirement applies to BWR's only.

A00144G2 A1.68-1

B/B-FSAR AMENDMENT 46 JANUARY 1985 l

Appendix A.5.j states " Verify that plant performance is as expected for rod runback and partial scram."

The Applicant asserts _that these particular events are applicable to BWRs only, and therefore will not be performed at Byron.

Appendix A.S.gg states "If appropriate for the facility design, conduct tests to determine operability of equipment provided for anticipated transient without scram (ATWS), if not previously done (254)."

The Applicant does not intend to perform this testing because current facility design does not include special ATWS equipment.

Appendix A.S.kk states " Demonstrate that the dynamic response of the plant is in accordance with design for the loss of or bypassing of the feedwater heater (s) from a credible single failure or operator error that would result in the most severe case of feedwater temperature reduction (504, 904)."

The Applicant asserts that secondary side design assures that there is no credible single failure or operator error that would cause the loss of or bypassing of the feedwater heaters. There-fore, this postulated transient does not require testing.

Appendix A.5.mm states " Demonstrate that the dynamic response of the plant is in accordance with design for the case of auto-matic closure of all main steam line isolation valves. For PWRs, justification for conducting the test at a lower power level, while still demonstrating proper plant response to this transient, may be submitted for NCR staf f review (1004) ."

The Applicant does not intend to perform this test because the closure of all main steam isolation valves will result in a turbine trip per Byron FSAR Subsection 15.2.4. The turbine trip test l will be performed at 100% power and is a more severe transient.

. The performance of the MSIV test would be redundant and would provide no additicnal information regarding plant response or capability. In effect, performance of this test would only result in unnecessary cycling of this equipment.

A test program has been established to ensure that all structures, systems, and components will satisfactorily perform their safety-related functions. This test program provides additional assurance that the plant has been properly designed and constructed and is ready to operate in a manner that will not endanger the health

! and safety of the public, that the procedures for operating the plant safely have been evaluated and have been demonstrated, and that the plant and procedures are fully prepared to operate

{ the facility in a safe manner.

A1.68-2 A0014463

B/B-FSAR AMENDMENT 46 JANUARY 1985 l

The test program includes simulation of equipment failures and control system malfunctions that could reasonably be expected to occur during the plant lifetime. The test program also in-cludes testing for interactions such as the performance of inter-lock circuits in the reactor protection system. It also deter-mines that proper permissive and prohibit functions are performed.

Care is taken to ensure that redundant channels of the equipment are tested independently.

The initial startup testing, conducted af ter the fuel loading and before commercial operation, will confirm the design bases and demonstrate, where practical, that the plant is capable of withstanding the anticipated transient and postulated accidents.

A detailed description of the test program is provided in FSAR Chapter 14.0.

I i

i l

\

l A1.68-3 A30144G4 l

- - . . . - - - . _ - . ~ _ _ - - - .

.= . . _ .

. . l l

l B/B-FSAR AMENDMENT 37 MARCH 1982 REGULATORY GUIDE 1.68.1 Revision 1, January 1977 PREOPERATIONAL AND INITIAL STARTUP TESTING OF FEEDWATER AND CONDENSATE SYSTEMS FOR BOILING WATER REACTOR POWER PLANTS This guide is pertinent to BWR's only.

l I 3

i A0014465 ,

1 A1.68.1-1

B/B-FSAR AMENDMENT 45 JUNE 1984 REGULATORY GUIDE 1.68.2 Revision 1, July 1978 INITIAL STARTUP TEST PROGRAM TO DEMONSTRATE REMOTE SHUTDOWN CAPABILITY FOR WATER-COOLED NUCLEAR POWER PLANTS l The Applicant complies with the position of this regulatory guide.

Refer to Table 14.2-86.

i l

A1.68.2-1 A0014466

1 B/B-FSAR AMENDMENT 37 MARCH 19 82 i

REGULATORY GUIDE 1.69 Revision 0, December 1973 CONCRETE RADITION SHIELDS FOR NUCLEAR POWER PLANTS The Applicant complies with the Regulatory Position of this guide. Concrete radiation shielding is discussed in Subsection 12.3.1.5.

I r

1 l

I

! A003.4467 !

l t l-

. Al . 6 9 _. ,. _-_ . - - . ._

B/b-FSAR Af!ENDME:JT 37 MARCH 19 82 REGULA10RY GUIDE 1.70 Revision 2, September 1975 STANDARD FORMAT' AND CONTENI OF SAFETY ANALYSIS REPORIS FOR NUCLEAR POWER PLANTS This FSAR is written in accordance with tne content and formdt set forth oy Regulatory Guide 1.70, Revision 2.

J 0

a f

i A303A4Gs A1.70-1

B/E-FSAR AMENDMENT 37 MARCH 1982 REGULAICRY GUIDE 1.71 Revision 0, December 1973 h ELDER QUALIFICATION F OR AREAS OF LI'41TED ACCESS 1b1LITY The Applicant maintains that limited-accessibility qualitication or requalification, exceeds ASME Section III and 1A requirements and is an unduly restrictive and unnecessary requirement.

Acceptaoility of welds will be determined by required examinations. Multiple production welds of similar components in the shop will oe subjected to close control and supervision achieving tne same purpose as the Guide. (See Paragraphs 5.3.1.4, 7.5.2.4, and 10. 3. 6. 2 tor further information.)

I i

l l

l i

I i

l A003.44G3 A1.71-1

B/B-FSAR AMEND!!ENT 37 MARCII 1982 REGULATORY GUIDE 1.72 l

l Revision 1, February 1978 i

SPRAY POND PLASTIC PIPING Inis regulatory guide does not apply to this application, since the Byron /Braidwood design does not utilize spray ponds.

1 i

ASP 14470 A1.72-1 i

l

B/B-FSAR AMENDMENT 37 MARCH 1982 l

REGULATORY GUIDE 1.73 Revision 0, January 1974 QUALIFICATION TESTS OF ELECTRIC VALVE OPERATORS INSTALLED INSIDE THE CONTAINMENT OF NUCLEAR POWER PLANTS This Regulatory Guide indicates the NRC acceptance (with certain qualifications) of the requirements of IEEE 382-1972, "IEEE Trial-Use Guide for Type Test of Class I Electric Valve Operators for Nuclear Power Generating Stations."

The Applicant complies with the objectives set forth in this regulatory guide as indicated in Subsections 6.2.42 and 8.1.13.

A1.73-1  !

A0014471 l l

B/B-FSAR AMENDMENT 37 MARCH 1982 REGULATORY GUIDE 1.74 Revision 0, February 1974 t

1 QUALITY ASSURANCE TERMS AND DEFINITIONS Regulatory Guide 1.74 identifies quality assurance terms and acceptable definitions that are important to the under-standing of the quality assurance requirements'for design, construction and operation of nuclear power plant structures, systems, and components.

The Applicant complies with this regulatory guide. Refer to Chapter 17.0 for further information on the Commonwealth Edison Quality Assurance Program.

i

~0014472 A

A1.74-1

. _ _ . _ = - - = - _ .. - . _ - - - _ _ - . _ _ _. _ __

B/B-FSAR AMENDMENT 37 MARCH 1982 i

REGULATORY GUIDE 1.75 Revision 2, September 1978

. PHYSICAL INDEPENDENCE OF ELECTRIC SYSTEMS f

NSSS Scope The commitment to comply with the intent of the requirements of this guide by the Applicant is present in Subsections 7.1.2.2.1 and 7.1.2.2.2.

f Non-NSSS Scope Physical Independence of Redundant Electric Systems is dis-j cussed in Subsections 8.1.14 and 8.3.1.4.  ;

The Applicant complies with the requirements of this guide with the exceptions and/or clarifications to the Regulatory Positions identified and justified below:

1 Regulatory Position Cl Section 3, " Isolation Device" should be supplemented as follows: "(Interrupting devices actuated only by 3

fault current are not considered to be isolation devices within the context of this document.)"

i Applicant's Position Interrupting devices actuated only by fault current may

be used as isolation devices provided that the coordi- '

nation criteria of IEEE 384-1977, Section 6.1.2 are met.

! Justification of Applicant's Position  ;

i' There is no technical justification for precluding the use of Class lE circuit breakers actuated only by fault i or overload current as a circuit interrupting or isolation l' device. (For further discussion of this subject, see S&L letter to the NRC dated December 21, 1978, NRC's response dated March 28, 1979, and Question 040.73.) -

l Byron /Braidwood Design Although the Applicant believes that a single circuit  !

breaker (actuated by fault current only) provides adequate  !

isolation, the Byron /Braidwood design will incorporate i the following additional features to further ensure  !

l isolation and thus satisfying NRC concerns.  !

[

A1.75-1 1 A0014473  !

O

• B/B-FSAR AMENDMENT 37 MARCH 1982 t

The Applicant (where practical) will provide two inter-rupting devices (in series) actuated only by fault current.

These two interrupting devices will be: 1) Class lE qualified; 2) will be coordinated with their upstream interrupting device; and 3) will be periodically tested to verify coordination. Any remaining non-Class lE loads (not utilizing two interrupting devices) will be tripped from the Class lE buses with a Safety Injection coincident with loss of offsite power signal. The cables which supply non-Clss lE loads from redundant Class lE buses are routed through separate raceways.

Regulatory Position C2 Section 3, " Raceway". Interlocked armor enclosing cable l should not be construed as a " raceway." '

i l

I I

I A1.75-la A W14474

B/B-FSAR AMENDMENT 35 DECEMBER 1981 Applicant's Position Although not a " raceway" in the same sense as a conduit or cable tray, recognition of, and design credit for the additional protection provided by the metallic jacket of interlocked armored cable should be included in the Regulatory Guide. Use of armored cable, in lieu of the separation distances stated in the Regulatory Guide, should be permitted when justified by specific testing and/or analysis, as provid-ing the required degree of protection for Class 1E circuits against specific credible hazards.

Justification of Applicant's Position There is no technical justificaiton for precluding the use of armored cable, in lieu of separation distances, to provide adequate isolation between Class lE and non-Class lE circuits and between redundant Class lE circuits, when shown to be adequate by specific testing and/or analysis.

Regulatory Position C6 Analyses performed in accordance with Sections 4.5(3), 4.6.2, and 5.1.1.2 should be submitted as part of the Safety Analysis Report and should identify those circuits installed in accor-dance with these sections.

Applicant's Position The referenced analysis, when performed to justify deviation from specific requirements of standard IEEE 384-1974, shall be prepared on a case-by-case basis, shall be documented and be on permanent file, available for NRC review, but will not be an integral part of the Safety Analysis Report.

j Justification of Applicant's Position The Applicant's posit.on is consistent with that taken for l other plant design records; e.g., routine design calculations, design document revisions, etc.

Regulatory Position C7 Non-Class lE instrumentation and control circuits should not be exempted from the provisions of Section 4.6.2.

l Applicant's Position Low energy non-Class lE instrumentation and control circuits are not required to be physically separated or electrically l isolated from " associated" circuits provided (a) the non-l Class lE circuits are not routed with " associated" circuits A0014475 A1.75-2

B/B-FSAR AMENDMENT 43 SEPTEMBER 1983 5

of a redundant division, and (b) they are analyzed to demon-strate that Class lE circuits are not degraded below an acceptable level. As part of the analysis, consideration  ;

shall be given to the potential energy and identification l of the circuits involved.

Justification of Applicant's Position The Applicant's position is consistent with the industry consensus position regarding required separation between non-Class lE circuits and " associated" circuits, taken in the 1977 and 1981 revisions to IEEE-384, Section 4.6.l(4) .

Regulatory Position C8 Section 5.1.1.1 should not be construed to imply that adequate separation of redundant circuits can be achieved within a confined space such as a cable tunnel that is effectively unventilated, l

Applicant's Position Adequate separation of redundant Class lE circuits can be achieved in areas of the plant that are effectively unventi-lated.

Justification of Applicant's Position There is no technical justification for precluding the routing

of redundant Class lE circuits through areas of the plant i that may be " effectively unventilated" provided that adequate physical separation is provided between redundant circuits i and appropriate thermal derating f actors for such circuits have been incorporated into the plant design.

I Regulatory Position C9

' Section 5.1.1.3 should be supplemented as follows:

"(4) Cable splices, in raceways, should be prohibited."

Applicant's Position Cable splices, either within -raceways or at the interface of raceways and equipment, etc., are permitted provided they are intended by the plant design as indicated on the design documents.

Justification of Applicant's Position There is no technical justification for precluding the use f of cable splices within raceways or at their interfaces with equipment, etc., provided that they are an integral ,

l part of the plant design as indicated on the design documents.

A1.75-3 AOO3A4 O

B/B-FSAR AMENDMENT 43 SEPTEMBER 1983 Regulatory Position C10 Section 3.1.2, the phrase "at a sufficient number of points" should be understood to mean at intervals notAlso, to exceedthe pre-5 feet throughout the entire cable length.

' f erred method of marking cable is color coding.

l ,

Applicant's Position Cable installed in exposed Class lE and " associated" circuit raceways shall be identified in a manner of sufficient dura-bility and at sufficient intervals to facilitate initial verification that the installation is in conformance with the separation criteria. Methods of providing the justifi-i cation, other,than color coding of the cable jacket, are acceptable.

I Justification of Applicant's Position 3

There is no technical justification for requiring the inter-vals of identification of such cables to not exceed Cable every system 5 feet throughout the entire cable length. r designs employing less frequent identification intervals and that provide for verification that the installation is in i

conformance with the separation criteria are acceptable.

Use of cable jacket color coding alone, as a method of provid-ing cable identification, may not beusing as effective as alterna-unique cable identi-l tive methods. Other methods, e.g.,

i fication number with a segregation code, could be a more positive method of facilitating verification that the cable installation is in accordance with the design separation requirements.

i Regulatory Position Cll Section 5.1.2 should be supplemented as follows: "The method of identification used should be simple and should preclude l i

the need to consult any reference material to distinguish between Class lE and non-Class lE circuits, between non-l Class lE circuits ' associated' with different redundant

Class lE systems, and between redundant Class lE systems."

! Applicant's Position I

The method of initial installation verification need not preclude consultation of reference documents.

Justification of Applicant's Position f

There is no technical justification for precluding use of reference documents during the installation verification A1.75-4 AW14477

a . .

B/B-FSAR . AMENDMENT 43 SEPTEMBER 1983 process (e.g., use of design documents and installation records). A system based upon the use of such reference documents could be a most effective check that a cable is installed in accordance with the design documents and in a raceway of compatible segregation assignment.

Regulatory Position C12 Pending issuance of other acceptable criteria, those port ons i of Section 5.1.3 (exclusive of the Note following the second .

paragraph) that permit the routing of cables through the I cable spreading area (s) and, by implication, the control Also, Section room, should not be construed as acceptable.

5.1.3 should be supplemented as follows: " Where feasible, redundant cable spreading areas should be utilized."

Applicant's Position 4

Power cables installed in dedicated solidly enclosed metallic raceways in air (e.g., rigid steel conduit or solid cable trays with solid flush covers), may be routed through those

' areas designated as " cable spreading areas," where justified by analysis or other suitable means.

Justification of Applicant's Position There is no technical justification to preclude the routing of power cables through cable spreading areas when they are installed in such a manner to present no hazard to other cabling, generally of a lower energy level, within the area.

Regulatory Position Cl4 i

Section 5.2.1 should be supplemented as follows

"And should have independent air supplies."

Applicant's Position Redundant standby generating units shall be placed in separate

safety class structures and shall be provided with separate

' ventilation and combustion air systems.

Justification of Applicant's Position The Applicant's position is an interpretation of what is believed to be the intent of the Regulatory Position.

Regulatory Position C15 Where ventilation is required, the separate safety class structures required by Section 5.3.1 should be served by independent ventilation systems.

^1' 5 A0014478 3 mm- t -- w ,,e-,,,-- -w-m - . - - - - ,

4w --,e.,, ,,s. > , , ,,- ,,,,-g ,--..~_--.,,,--.-w , -

i B/B-FSAR AMENDMENT 35 j DECEMBER 1981 i Applicant's Position Redundant batteries shall be housed in separate safety class structures, i.e., separate from one another, not necessarily separate from everything else within its own safety division.

For example, a battery may be placed in the same safety class structure as the switchgear for that division.

Justification of Applicant's Position The Applicant's position is an interpretation of what is believed to be the intent of the Regulatory Position.

Regulatory Position C17 Regulatory Guide Position on Section 4.6.1 " Separation from Class lE Circuits," of IEEE Std 384 (1974)

By not modifying Section 4.6.1 of IEEE Std 384 (1974) in a Regulatory Position, the Regulatory Guide has endorsed it as stated in the IEEE standard.

Applicant's Position There is no justification for precluding the use of techni-cally acceptable analysis to justify, on a case-by-case basis, exceptions to the generally stated criteria for separa-tion of non-Class lE circuits, from Class lE circuits. When such analysis demonstrates that the following requirements are met, the non-Class lE circuits involved need not be classified as " associated" circuits.

?

For specific cases, where cable termination or routing arrange-ments (e.g., cables leaving cable trays in free air entering n

equipment or passing through co'duit sleeves in walls) limit the available separation distances between non-Class lE and Class lE cables, to less than the minimum separation applicable to redundant cables in raceways, such lesser separations are permitted provided that a documented analysis is performed to demonstrate that:

, a. the non-Class lE circuits are not routed with Class lE circuits of a redundant division or circuits " associated" with a redundant division, and

b. the Class lE circuits involved are not degraded below and acceptable level.

The analysis will include consideration of the potential energies of the circuits involved; the physical and electrical isolation (i.e., barriers) provided for the circuits by the Al.75-6 A

l l

B/B-FSAR AMENDMENT 35 DECEMBER 1981 l

cable insulation, the cable shielding, and the cable jacket- 1 ing systems; the degree of environmental qualification and fire retardant characteristics of the cables; and the poten-tial for hazards in the specific area involved.

Justification of Applicant's Position The Applicant's position is consistent with the industry con-sensus technical position stated in the 1977 revision to IEEE-384, Section 4.6.l(3).

t l

i l

l l

A1.75-7 A0014460 L

B/B-FSAR AMENDMENT 37 MARCH 1982 REGULATORY GUIDE 1.76 Revision 0, April 1974 DESIGN-BASIS TORNADO FOR NUCLEAR POWER PLANTS The plant design conforms to the regulatory position as de-scribed in Section 3.3.

l i

l l

A1.76-1 A0014481 l I

b/3-FSAR AMENDMENT 37 MARCH 1982 REGULATORY GUIDE 1.77 i l

Revision 0, May 1974 ASdUMPa'10N5 USED FOR EVALUATING A CONTROL ROD EJECTION ACCILEh1 rOR PRESSURIZED WATER REACTORS (Refer to S uosections 15. 4. 7 and 15.4.8.3 for details of this a nalysi s. ) .

nestinghouse methods and criteria are documented in WCAP-7586 Revision 1A which has been reviewed and accepted by the NRC.

The results of their analyses show compliance with the Regulatory Position given in Sections C1 and C3 of Regulatory Guide 1.77.

dowever, they take exception to Regulatory Guide Position C2 which implies that the rod ejection accident should be considered as a emergency condition. Westinghouse considers this a faulted condition as stated in ANSI N16.2, " Nuclear Safety Criteria for the Design or Stationary Pressurized Water Reactor Plants".

Faulted condition stress limits will os applied for this accident.

l A1. 77- 1

B/B-FSAR AMENDMENT 37 MARCH 19 82

.- REGULATORY GUIDE 1.78 Revision 0, June 1974 ASSUMPTIONS FOR EVALUATING THE HABITABILITY OF A NUCLEAR POWER PLANT CONTROL ROOM DURING A POSTULATED HAZARDOUS CHEMICAL RELEASE The Applicant complies with this Regulatory Guide. Refer to Section 2.2 and Subsection 6.4.1 for further information.

I

\

A M 44S3

^

l A1.78-1

B/B-FSAR AMENDfiENT 37 tiARCH 1982 REGULATORY GUIDE 1.79 Revision 1, September 1975 PREOPERATIONAL TESTING OF EMERGENCY CORE COOLING SYSTEMS FOR PRESSURIZED WATER REACTORS The Applicant complies with the requirements of this guide.

The containment spray system will be tested up to the con-tainment isolation valve while taking suction from the refueling water storage tank. In testing the RHR system under the recirculation conditions, the containment sumps will be filled with cold water up to print elevation 376 feet 9 inches. The adequacy of NPSH available under postaccident recirculation conditions to the RHR pumps will be corrected for water elevation, temperature, and run out flow through the containment spray pumps. See Chapter 14.0 for further discussion of preoperational testing.

A1.79-1 AE4E

B/B-FSAR AMEND!iENT 37 ;

MARCH 1982 i REGULATORY GUIDE 1.80 l

Revision 0, June 1974 PRcOPERATIONAL TESTING OF INSTRUMENT AIR SYSTEMS The air systems in tne Byron /Braidwood design are designated Satety Category II, Quality Group D. As non-safety related e quipment, the air system does not come under the provision of this guide.

l l

l A1.60-1 AW144SS

B/3-FSAR AMENDMENT 37 MARCH 1982 REGULATORY GUIDE 1.81 Revision 1, January 1975 l

SHARED EMERGENCY AND SHUTDOWN ELECTRIC SYSTEMS FOR MULTIUNIT NUCLEAR POWER PLANTS The Byron /Braidwood design complies with the requirements of this regulatory guide (which indicates the acceptable methods of compliance with General Design Criterion 5). The independence of each unit's onsite electrical systems is further discussed in Subsection 8.1.15.

l A1.81-1 -

l A303.4486 l

2 B/B-FSAR AMENDMENT 44 DECEMBER 1983 REGULATORY GUIDE 1.82 Revision 0, June 1974 SUMPS FOR EMERGENCY CORE COOLING AND CONTAINMENT SPRAY SYSTEMS The Applicant complies with the Regulatory Position of this guide as discussed below:

1) "A minimum of two sumps should be provided, each with sufficient capacity to serve one of the redundant halves of the ECCS and CSS systems."

Two sumps are provided per unit; each having sufficient capacity to serve one of the redundant halves of the ECC and CSS system.

2) "The redundant sumps should be physically separated from each other and from high energy systems by structural barriers, to the extent practical to preclude damage to the sump intake filters by whipping pipes or high-velocity jets of water or steam."

The redundant sumps are located approximately 15 feet apart and are physically separated. No high energy lines are located within 12 feet of the sumps. This precludes damage to the sump intake filters by whipping pipes or high-velocity jets of water or steam.

I

3) "The sumps should be located on the lowest floor eleva-tion in the containment exclusive of the reactor cavity.

At a minimum, the sump intake should be protected by two screer.s: (1) an outer trash rack, and, (2) a fine inner screen. The sump screens should not be depressed below the floor elevation."

The sumps are located on elevation 377 feet in the con-tainment, the lowest floor elevation in the containment exclusive of the reactor cavity. Each sump intake is protected by three screens. The two outer screens are l located above elevation 377 feet.

4) "The floor level in the vicinity of the coolant sump location should slope gradually down away from the sumps."

As stated in Appendix A, the water level in the contain-I ment at the end of safety injection will be 5 feet above the floor level. Sloping the floor would provide little protection against deoris at these levels. Redundant A1.82-1 A W3.4487 l -

B/B-FSAR AMENDMENT 44 DECEMBER 1983 outer screens have been provided at each sump. If one outer screen is totally blocked by debris, the other will still emit water into the sump.

5) "All drains from the upper regions of the reactor building should terminate in such a manner that direct streams of water, which may contain entrained debris, will not impinge on the filter assemblies."

The filters are located such that direct streams of water which may contain entrained debris will not impinge on the filter assemblies, j 6) "A vertically mounted outer trash rack should be provided to prevent larger debris from reaching the fine inner screen. The strengths of the trash rack should be con-t sidered in protecting the inner screen from missiles and large debris."

The outer screen on the recirculation sumps is 3/8 inch square wire mesh.

7) "A vertically mounted fine inner screen should be provided.

The design coolant velocity at the inner screen should be approximately 6 cm/sec (0.2 ft/sec). The available surface area used in determining the design coolant j velocity should be based on one-half of the free surface i area of the fine inner screen to conservatively account for partial blockage. Only the vertical screens should be considered in determining available surface area." l The outer screen on the sumps has been designed to meet the criteria of a maximum coolant velocity of 0.2 ft/sec with one-half of the free surface area blocked. This approach has been taken because the sump geometry does not allow for an inner screen of sufficient size to achieve this low flow velocity.

8) "A solid top deck is preferable, and the top deck should be designed to be fully submerged after a LOCA and com-1 pletion of the safety injection."

l The top deck is 1/4 inch stainless steel checkered plate.

l 9) "The trash rack and screens should be designed to withstand l the vibration motion of seismic events without loss

' of structural integrity."

A1.82-2 A00.4488 1

i B/B-FSAR AMENDMENT 44

' DECEMBER 1983 All of the screen mountings and the sump itself are Category I and are designed to withstand an SSE event,

10) "The size of openings in the fine screen should be based on the minimum restrictions found in systems served by the sump. The minimum restriction should take into account the

' ,; overall operability of the system served."

i ,

The restriction upon the containment spray system is the particles be less than 1000 microns in size.

l The maximum particle size capable of passing through the fine vertical inner screen is less than 750 microns.

l l 11) " Pump intake locations in the sump should be carefully considered to prevent degrading effects such as vortexing on the pump performance."

4 i

The pump intake is located off the side of the sump near the bottom. This location should prevent degrading effects on the pump performance.

12) " Material for trash racks and screens should be selected to avoid degradation during periods of inactivity and operation and should have a low sensitivity to adverse effects such as stress-assisted corrosion that may be induced by the chemically reactive spray during LOCA conditions."

i The screens on the recirculation sump are 316 stainless steel, t

l 13) "The trash rack and screen structure should include access openings to facilitate inspection of the structure and pump suction intake."

A manway has been provided for inspection of the sump internals.

14) " Inservice inspection requirements for coolant sump

components (trash racks, screens, and pump suction inlets) i should include the following:

a) Coolant sump components should be inspected during every refueling period down time, and, l

b) The inspection should be a visual examination of the components for evidence of structural distress of Corrosion."

This requirement will be adhered to.

f A1.82-3 i AWM489

B/B-FSAR AMENDMENT.37 MARCH 1982 outer screen that will encompass both of the existing recirculation sumps. The outer screen area will be sized such that the water inlet velocity will be 0.2 ft/sec with one-half of the screen area blocked.

The screen design will be similar to the existing outer screens with openings on all four sides and a solid metal top. The screen will be elevated on a 6- to 12-inch high concrete curb which will allow debris to be collected and not block the screen opening. This curb will have the same effect as a gradual sloping of the floor away from the screen.

The applicant believes that with this modification along with the existing design, the containment recirculation sumps meet the intent of Regulatory Guide 1.82.

I i

l l

A1.82-4 A003A490

B/B-E3AR AMENDMENT 37 MARCH 1982 REGULATORY JUIDE 1.83 Revision 1, July 1975 INdEAVICE IN6PECTION OF PRESSURIst0 WATER EEACTOR STEAf4 GENERATOR _AUBES segulatory Juide 1.83 descrioes an acceptable metrod of complying witn tha commission's regulations with regard to inservice inspection or pressurized hater reuctor steam generator tubec.

The plant design incluaes the features or Regulatory Position C.1. Iechnical specitications zor assuring inspection and reporting as recommended Ly segulatory Positions C. 2 tnrougn C.9 are incorporated in Chapter 16.0.

i A1.63-1 ffj(q,y91

B/B-FSAR AMENDMENT 37 MARCH 1982 f

REGULATORY GUIDE 1.84 i l I Revision 18, August 1981 DESIGN AND FABRICATION CODE CASE ACCEPTABILITY ASME SECTION III DIVISION 1 l'

The applicant complies with the Regulatory Position. All Code Cases on the approved list as endorsed by the regulatory guide i are acceptable for application upon submittal of request by vendor. Code Case acceptability is limited to the requirements as specified in the " Inquiry" and " Reply" sections of each individual Code case.

Long-lead time components for the Byron /Braidwood project were ordered prior to the original effective dates for Regulatory Guides 1.84 and 1.85. Nevertheless, there are no known examples of Code Cases being applied to components of this project except those approved by either Regulatory Guides 1.84 or 1.85, with the following exceptions or special considerations:

i a. Code Case 1637 - The code case was used for the i purchase of heat oxchanger tubing. Authorization

for its use was obtained from NRC.

i

b. Code Case 1528 - Fracture toughness information for this code used in the construction of the steam generators and pressurizers has been supplied to the NRC by WCAP-9292, March 1978, " Dynamic Fracture Toughness of ASME SA-508 Class 2a and ASME SA-533 Grade A Class 2 Base and Heat Affected Zone Material and Applicable Weld Metals."

l (Refer to Subsection 5.4.2 for further information.)

In addition the following code cases have been approved by the NRC

a. N-272 Compiling Data Report Records

b. N-275 Repair of Welds

c. N-292 Depositing Weld Metal Prior to Preparing ,

ends for welding

d. N-295 Use of Previously Produced Material

! A1.84-1 l

A00244S2

B/B-FSAR AMENDMENT 37 MARCH 1982 REGULATORY GUIDE 1.85 ,

Revision 18, August 1981 MATERIALS CODE ACCEPTABILITY ASME SECTION III DIVISION 1 The applicant complies with the Regulatory Position. All Code Cases on the approved list as endorsed by the regulatory guide are acceptable for application upon submittal of request by vendor. Code Case acceptability is limited to the requirements as specified in the " Inquiry" and " Reply" sections of each individual dual Code Case.

Long-lead time components fo, the Byron /Braidwood project were ordered prior to the original effective dates for Regulatory Guides 1.84 and 1.85. Nevertheless, there are no known examples of Code Cases being applied to components of this project except those approved by either Regulatory Guides 1.84 or 1.85, with the following exceptions or special considerations

a. Code Case 1637 - The code case was used for the purchase of heat exchanger tubing. Authorization for its use was obtained from NRC.

b. Code Case 1528 - Fracture toughness information for this code case used in the construction of the steam generators and pressurizers has been supplied to the NRC by WCAP-9292, March 1978, " Dynamic Fracture Toughness of ASME SA-508 Class 2a and ASME SA-533 Grade A Class 2 Base and Heat Affected Zone Material and Applicable Weld Metals."

(Refer to Subsection 5.4.2 for further information.) '

l

+

l l

l Al.85-1 AOOJ..M'i3

_-_. _ _ _ _ _ _ - _ _ .. a

B/B-F3AR AMEND 11ENT 37 MARCH 1982 REGULATORY GUIDE 1. 86 Revision 0, June 1974 TERMINATION OF OPERATING LICc,NSES FOR NUCLEAR Rt. ACTORS Regulatory Guice 1. 86 will be complied with by the Regulatory Staff rar the termination of operating licenses at tne end of the station design life.

i i

l 1

A0"J.44S4 A1.86-1 l - . - . . _- _ " we+,.,.,_ __

L/B-FSAR A!!ENDMENT 37 11 ARCH 1982 REGULATORY GUIDE 1.87 Revision 1, June 1975 GUIDANCE FOF CONSTRUCTION OF CLASS 1 CO:(PONENTS IN ELh.VATED-TEMPERATURE REACTORS This guioe is not pertinent to this application, since the

myron/Eraidwood reactors are not "nigh-temperature reactors. "

i l

I A0014495 A1.87-1

. o

. o B/B-FSAR AMENDMENT 45 JUNE 1984 REGULATORY GUIDE 1.88 Revision 2, October 1976 COLLECTION, STORAGE, AND MAINTENANCE OF NUCLEAR POWER PLANT QUALITY ASSURANCE RECORDS The Applicant complies with this regulatory guide with the comment noted below. Refer to Chapter 17.0 of the FSAR for further information on the Commonwealth Edison Quality Assurance Program.

The Applicant complies with Section 17.1, paragraph II.17.4 of the Standard Review Plan (NUREG 75/087, Rev.1) for the temporary record storage facilities used during the construc-tion period. The permanent record storage facilities will meet the recommendations of ANSI N45.2.9-1974, which is referenced in this regulatory guide, with the exception of Item 5.6 (3) doors. Frames and hardware should be Class A fire-rated with a recommended 4-hour rating. The doors at Byron and Braidwood are UL listed 3-hour fire-rated hollow metal doors due to the fact the industry has not performed 4-hour fire tests on these doors.

1 A00.1.44SG A1.88-1

B/B-FSAR AMENDMENT 37 MARCH 1982 REGULATORY GUIDE 1.89 Revision 0, November 1974 QUALIFICATION OF CLASS lE EQUIPMENT FOR NUCLEAR POWER PLANTS NSSS Scope For Westinghouse NSSS Class lE equipment, Westinghouse will meet IEEE-323-1974, "IEEE Standard for Qualifying Class lE Equipment for Nuclear Power Generating Stations" including the Nuclear Power Engineering Committee (NPEC) Position Statement of July 24, 1975, and Regulatory Guide 1.89, by an appropriate combination of any or all of the following: type testing, operating experience, qualification by analysis, and on-going qualification. This commitment will be satisfied by implementa-tion of the final approved version of WCAP-8587.

Non-NSSS Scope The extent of the Applicant's commitment to comply with the requirements of Regulatory Guide 1.89 is presented in Subsections 3.11.2 and 8.1.16.

l A1.89-1 1

(

AGU3A4S7

B/B-FSAR AMENDoiENT 37 4 MARCH 1982 REGULATORY GUIDE 1.90 Revision 1, August 1977 INSERVICE INSPECTION OF PRESTRESSED CONCRETE CONTAINMENT STRUCTURES WITH GROUTED TENDONS The containment design does not use grouted tendons. Thus, this guide is not applicable to Byron /Braidwood.

f 1

i i

i i

r A003A498 A1.90-1

B/B-FSAR AMENDMENT 37 MARCH 1982 l

REGULATORY GUIDE 1.91 l Revision 1, February 1978 EVALUATION OF EXPLOSIONS POSTULATED TO OCCUR ON TRANSPORTATION ROUTES NEAR NUCLEAR POWER PLANTS The plant design conforms to the regulatory position as described in Subsection 2.2.3.2.

i l

l A1.91-1 l A0014499

B/B-FSAR AMENDMENT 37 MARCH 1982 REGULATORY GUIDE 1. 92 Revision 1, February 1976 COMBINING MODAL RESPONSES AND SPATIAL COMPONENTS IN SEISMIC RESPONSE ANALYSIS The plant design conforms to the regulatory positions as de-scribed in Subsections 3.7.2 and 3.7.3.7 with one exception:

the absolute value of the response was not used in the double sum method. The discussion and results presented by Rosenblueth (Reference 1) and Singh (Reference 2) justify the use of the algebraic sign when combining closely spaced modes using the double sum method.

References

1. E. Rosenblueth and J. Elorduy, " Response of Linear Systems to Certain Transient Disturbances," Proceedings, Fourth World Conference on Earthquake Engineering, Vol. 1, Santiago, Chile, 1969.

2. A. K. Singh, S. L. Chu, and S. Singh, " Influence of Closely Spaced Modes in the Response Spectrum Method of Analysis,"

ASCE Specialty Conference on Structural Design of Nuclear Power Plant Facilities, Chicago, Illinois, December 17-18, 1973.

I A1.92-1  !

A0014SCO l

1

B/B-FSAR AMENDMENT 37 MARCH 1982 REGULATORY GUIDE 1.93 Revision 0, December 1974 AVAILABILITY OF ELECTRIC POWER SOURCES Availability of electric power sources is discussed in Subsection 16.3/4.8.

The Applicant complies with the requirements of this guide with the following exception:

Regulatory Positions C.1, C.2, and C.4 refer to a 72-hour time interval for power operation when the available power sources are less than the " Limiting Conditions for Operation."

The Applicant uses a 7-day time interval in place of the 72-hour time interval contained in this guide.

A003A501 Al.93-1

~~~~~ ~

B/B-FSAR AMENDMENT 38 MAY 1982 l

REGULATORY GUIDE 1.94 Revision 1, April 1976 QUALITY ASSURANCE REQUIREMENTS FOR INSTALLATION, INSPECTION, AND TESTING OF STRUCTURAL CONCRETE AND STRUCTURAL STEEL DURING THE CONSTRUCTION PHASE OF NUCLEAR POWER PLANTS This Applicant complies with this regulatory guide. Refer i to Appendix B of the FSAR and the response to Question 421.19 l for further information.

a A1.94-1 AI.)O24502

B/B-FSAR AMENDMENT 37 MARCH 1982 REGULATORY GUIDE 1.95 Revision 1, January 1977 PROTECTION OF NUCLEAR POWER PLANT CONTROL ROOM OPERATORS AGAINST AN ACCIDENTAL CHLORINE RELEASE The Applicant complies with the requirements of this guide, as discussed in FSAR Subsection 6.4.4.

l h

l A0014503 A1.95-1 l

B/B-FSAR AMENDMENT 37 MARCH 1982 REGULATORY GUIDE 1.96 I Revision 1, June 1976 DESIGN OF MAIN STEAM ISOLATION VALVE LEAKAGE CONTROL SYSTEMS FOR BOILING WATER REACTOR NUCLEAR POWER PLANTS The requirements of this guide are not applicable to pressurized water reactors.

i A005.4504 A1.96-1

B/B-FSAR MAY 1983 REGULATORY GUIDE 1.97 Revis'.on 2, December 1980 INSTRUMENTATION FOR LIGHT-WATER-COOLED NUCLEAR POWER PLANTS TO ASSESS PLANT COllDITIONS DURING AND FOLLOWING AN ACCIDENT The schedule for subnitting the Applicant's response to Regulatory Guide 1.97, Revision 2 is included in the April 14, 1983 corporate response to Generic Letter No. 82-33, " Supplement 1 to NUREG-0737 -

Requirements For Emergency Response Capabilities." Byron /Braidwood Project Engineering has completed a Regulatory Guide 1.97, Revision 2 Survey which will be submitted for staff review in May 1983.

A1.97-1 l A0014505

L/B-FSAR AMENDMENT 37 MARCH 1982 1

1 REGULATORY GUIDE 1.98

't Revision 0, March 1976 ASSUMPTIONS USED FOR EVALUATING THE POTENTIAL RADIOLOGICAL CONSEQUENCES OF A RADIOACTIVE OFFGAS SYSTEM FAILURE IN A BOILING WATER REACTOR The requirements of this guide are not applicable to pressurized water reactors.

1 i

A00145cc A1.98-1

B/B-FSAR AMENDMENT 37 MARCH 1982 REGULATORY GUIDE 1.99 Revision 1, April 1977 EFFECTS OF RESIDUAL ELEMENTS OR PREDICTED RADIATION DAMAGE TO REACTOR VESSEL MATERIALS Regulatory Guide 1.99 was issued after procurement of the Byron /Braidwood Nuclear Plant Units 1 and 2 reactor vessels.

However, the Byron Unit 1 reactor vessel meets the two positions of the guide as follows:

The residual elements in the vessel beltline material have been controlled to levels ruch that the predicted adjusted reference temperature at end-of-plant life, using the prediction methods given in the guide position C.1, will be less than 200' F and is therefore in agreement with the guide position C.2.

Although application of the guide procedures to the Byron Unit 1 reactor vessel results in the prediction that the reactor vessel material meets the end-of-life adjusted reference temperature criterion of the guide, the Westinghouse position is that the procedures set forth in the guide are overconservative at the higher fluences, and the restriction of the adjusted reference temperature at end of life to 200' F is technically unnecessary.

Details of the Westinghouse objections were transmitted to the staff via letter NS-CE-784 on September 22, 1975.

Although test data on the Byron Unit 2 and Braidwood Units 1 and 2 vessels is not available at this time, residual elements in the vessle beltline material have been controlled and it is expected that these vessels will also be in agreement with the guide position C.2.

Recent surveillance capsule data from Westinghouse plants indicate ND predicted by the guide is excessively that the shift in RT conservative. Referenbe letter NE-TMA-1843, T.M. Anderson (W) to S.J. Chilk (NRC) " Regulatory Guide 1.99," June 23, 1978.

Refer to FSAR Section 5.3 for further discussion.

A0014507 A1.99-1

1 I

B/B-PSAR AMENDMENT 37 MARCH 1982 REGULATORY GUIDE 1.100 i

Revision 1, August 1977 SEISMIC QUALIFICATION OF ELECTRIC EQUIPMENT FOR NUCLEAR POWER PLANTS NSSS Scope The Applicant is in compliance with the objectives of Regulatory Guide 1.100. The Westinghouse program for seismic qualification of safety-related electrical equipment is delineated in WCAP-8587, Revision 1. This program is currently under review by the NRC.

For further details, refer to Section 3.10.

Non-NSSS Scope The Applicant complies with the objectives of Regulatory Guide 1.100. The Applicant's approach to seismic cualification of Class lE equipment is discussed in Section 3.10.

l l

l l

l l

l l

l A1.100-1 '(0014508

B/B-FSAR AMENDMENT 38 MAY 1982 REGULATORY GUIDE 1.101 i Revision 2, October 1981 EMERGENCY PLANNING AND PREPAREDNESS FOR NUCLEAR POWER REACTORS The guidance provided by Regulatory Guide 1.101, Revision 2, was utilized in the preparation of the Applicant's emergency response plans. The Applicant complies with this regulatory guide as described in Section 13.3.

i .

l l

A1.101-1 1

s 1 l

i l

B/B-FSAR AMENDMENT 37 MARCH 1982 l

l REGULATORY GUIDE 1.102 Revision 1, September 1976 FLOOD PROTECTION FOR NUCLEAR POWER PLANTS The plant design conforms to the regulatory positions as de-scribed in Sections 2.4 and 3.4.

A1.102-1

'A0014510 i

j

^

. . 1 B/B-FSAR AMENDMENT 37 MARCH 1982 REGULATORY GUIDE 1.103 Revision 1, October 1976 POST-TENSIONED PRESTRESSING SYSTEMS FOR CONCRETE REACTOR VESSELS AND CONTAINMENTS This regulatory guide was withdrawn July 8, 1981, however, the plant design conforms to the regulatory positions as de-scribed in Subsection 3.8.1 and Appendix B.3.

l l

l l A1.103-1 1

A0014511

O '

B/B-FSAR AMENDMENT 37 MARCH 1982 REGULATORY GUIDE 1.104 Revision 0, February 1976 OVERHEAD CRANE HANDLING SYSTEMS FOR NUCLEAR POWER PLANTS Regulatory Guide 1.104 was withdrawn effective August 16, 1979.

A000A512 Al.104-1 I

l

B/B-FSAR RMENDMENT 37 tiARCH 1982 REGULATORY GUIDE 1.105 Revision 1, November 1976 l

INSTRUMENT SETPOINTS The Applicant complies with the regulatory position with the following exceptions keyed to paragraph numbers in the position:

Position c.5, requires locking devices on instrument set-point adjustment mechanisms. We have not specified such locking devices on our instrument data sheets, so that these would only be available to the extent that they are standard equipment. In general, locking devices are not required to maintain stable instrument setpoints and we believe that setpoint stability will not be improved by providing locking devices.

Position c.6, requires documentation of the assumptions used in selecting setpoint values and the margins between the setpoints and the limiting safety system values. The docu-mentation is to include definition of instrument setpoint drift rate and the relationship of the drift rate to testing intervals. The Byron /Braidwood design conforms to this

, position only to the degree that setpoints are documented on the instrument data sheets along with instrument range and the maxistum range of the parameter being measured with respect to the other requirements of Position c.6,, generic drif t rates are not generally available for any instruments since drift rates would be affected by the particular ser-vice to which the instrument was subjected. Testing inter-vals are set on the basis of past experience with the spe-cific instrument types in question.

l l

Al.105-1 A'3014513

. . l B/B-FSAR AMENDMENT 37 MARCH 1982 REGULATORY GUIDE 1.106

)

Revision 1, March 1977 THERMAL OVERLOAD PROTECTION FOR ELECTRIC MOTORS ON MOTOR-OPERATED VALVES The Applicant complies with the requirements of this Regulatory Guide. The Applicant has selected the method described in Regulatory Position C.2.

I l

A1.106-1 A0014514

B/B-FSAR AMENDMENT 37 MARCH 19 82 REGULATORY GUIDE 1.107 Revision 1, February 1977 QUALIFICATIONS FOR CEMENT GROUTING FOR PRESTRESSING TENDONS IN CONTAINMENT STRUCTURES The Byron /Braidwood containment design does not use grouted tendons, therefore this guide is not applicable to Byron /Braidwood.

l l

l l

I A 003..",53 5 A1.107-1

~~

B/B-FSAR AMENDMENT 41 FEBRUARY 1983 REGULATORY GUIDE 1.108 Revision 1, Auguct 1977 PERIODIC TESTING OF DIESEL GENERATOR USED AS ONSITE ELECTRIC POWER SYSTEMS AT NUCLEAR POWER PLAlff S The Applicant complies with the requirements of this guide, with the exception of the requirements in Regulatory Positions C . 2. a. 5 and C . 2.d.

The Applicant's position is justified by the briefing of Sep-tember 8, 1982 for the ACRS Subcommittee on AC/DC Power Systems, Recent Operating Experience With Emergency Diesel Generators.

The Applicant's position is based on the operating experience noted in the briefing and the conclusion reached is that reg-ulatory positions cited above actually decrease the reliability of the diesel generators.

l l

l A1.108-1 40034516

s . l B/B-FSAR AMENDMENT 38 MAY 1982 REGULATORY GUIDE 1.109 Revision 1, October 1977 CALCULATION OF ANNUAL DOSES TO MAN FROM ROUTINE RELEASES OF REACTOR EFFLUENTS FOR THE PURPOSE OF EVALUATING COMPLIANCE WITH 10 CFR PART 50, APPENDIX I The Applicant complies with the position of this regulatory guide as presented in Subsections 11.2.3.3 and 11.3.3.7.

l 1

A1.109-1

B/B-FSAR AMENDMENT 37 MARCH 1982 REGULATORY GUIDE 1.110 Revision 0, March 1976 COST BENEFIT ANALYSIS FOR RADWASTE SYSTEMS FOR LIGHT WATER-COOLED NUCLEAR POWER REACTORS The Applicant complies with the Annex to 10 CFR 50 Appendix I and therefore this guide is not applicable.

l l

l A1.110-1 ILS

B/B-FSAR A!!ENDMENT 37 fiARCH 19 82 REGULATORY GUIDE 1.111 f Revision 1, July 1977 METHODS FOR ESTIMATING ATMOSPHERIC TRANSPORT AND DISPERSION OF GASEOUS EFFLUENTS IN ROUTINE RELEASES FROM LIGHT WATER-COOLED REACTORS The Applicant complies with the requirements of this guide. l This is discussed further in Section 2.3.

l l

l l

A1.111-1 1

B/B-FSAR AMENDMENT 37 MARCH 19 82 REGULATORY GUIDE 1.112 Revision 0-R, May 1977 CALCULATION OF RELEASES OF RADIOACTIVE MATERIALS _

IN GASEOUS AND LIQUID EFFLUENTS FROM LIGHT-WATElt-COOLED POWER REACTORS The Applicant complies with the requirements of this guide.

This is discussed further in Section 11.2.

I i

i l

A*/.112-1 .4520

B/B-FSAR AMENDMENT 38 )

MAY 1982 l l

REGULATORY GUIDE 1.113 Revision 1, April 1977 i

ESTIMATING AQUATIC DISPERSION OF EFFLUENTS FROM ACCIDENTAL AND ROUTINE REACTOR RELEASES FROM THE PURPOSE OF IMPLEMENTING APPENDIX I The Applicant complies with the position of this regulatory guide as presented in Subsections 2.4.12, 2.4.13.3, and 11.2.3.

N l

s

's A0014521 A1.ll3-1

\

l B/B-FSAR A!!EMDMENT 37 liARCH 1982 REGULATORY GUIDE 1.114 l Revision 1, November 1976 GUIDANCE ON BEING OPERATOR AT THE CONTROLS OF A NUCLEAR POWER PLANT The Applicant complies with the requirements of this guide.

Refer to Subsection 13.1.2.2 for futher information.

A1.ll4-1 AOUS.4522

D/B-FSAR AMENDMENT 37 MARCH 1982 REGULATORY GUIDE 1.115 Revision 1, July 1977 PROTECTION AGAINST LOW TRAJECTORY TURBINE MISSILES The Applicant meets the objectives set forth in this Regulatory Guide as presented in Section 3.5 and Appendix C.

1 A1.ll5-1 A0014523

e B/B-FSAR AMENDMENT 43 SEPTEMBER 1983 REGULATORY GUIDE 1.116 a

Revision 0-R, June 1976 l

QUALITY ASSURANCE REQUIREMENTS FOR INSTALLATION, INSPECTION, AND TESTING OF MECHANICAL EQUIPMENT AND SYSTEMS The Applicant complies with this regulatory guide. Refer to Chapter 17.0 of the FSAR for further information on the Commonwealth Edison Company Quality Assurance Program.

9

\

l l

4 1

l

'I A1.ll6-1

B/B-FSAR AMENDMENT 37 MARCH 1982 REGULATORY GUIDE 1.117 i

Revision 1, April 1978 TORNADO DESIGN CLASSIFICATION This guide is applicable only to construction permit appli-cations docketed after May 30, 1978. The Byron /Braidwood construction permit application was docketed prior to this date. For a discussion of the Byron /Braidwood design, refer to FSAR Section 3.2.

A003.4525 A1.ll7-1

. . l l

l B/B-FSAR AMENDMENT 46 JANUARY 1985 REGULATORY GUIDE 1.118 )

Revision 2, June 1978 PERIODIC TESTING OF ELECTRIC POWER AND PROTECTION S'YSTEMS The Applicant complies with the requirements of this guide.

Refer to the Technical Specifications for information on l

periodic testing.

Al.ll8-1 A003.4526

1 l

B/B-FSAR AMENDMENT 37 11 ARCH 19 82 REGULATORY GUIDE 1.119 l

No Current Issue SURVEILLANCE PROGRAM FOR NEW FUEL ASSEMBLY DESIGNS This regulatory guide was withdrawn June 23, 1977.

l.

}

\

4 l-i l

A0014527 A1.119-1

B/B-FSAR AMENDMENT 37 MARCH 1982 REGULATORY GUIDE 1.120 Revision 1, November 1977 FIRE PROTECTION GUIDELINES FOR NUCLEAR POWER PLANTS The Applicant's positions on the regulatory positions are pro-vided in detail in Chapter 3.0 of the Fire Protection Report.

i i

k i

A1.120-1 1

A!JK!145' S

B/B-FSAR AMENDMENT 37

, MARCH 1982 REGULATORY GUIDE 1.121 i

Revision 0, August 1976 l l

BASES FOR PLUGGING DEGRADED PWR STEAM GNERATOR TUBES i

The Applicant complies with the Regulatory Position with the following comments and exceptions keyed to paragraph

numbers in the Position

Position C.1 Westinghouse interprets the term " unacceptable def'ects" to apply to those imperfections resulting from service induced mechanical or chemical degradation of the tube walls which i -have penetrated to a depth in excess of the Plugging Limit.

Position C.2.a(2) and C.2.a(4)

Westinghouse- has documented its opinions on Regulatory Guide 1.121 by corporate letter and has identified as the major exception the margin of 3 against tube tailure for normal i operation. Westinghouse defines tube failure as plastic 4

I deformation of a crack to the extent that the sides of the crack open to a non-parallel, elliptical configuration. The '

. tubing can sustain added internal pressure-beyond those values i' before reaching a condition of gross failure. We have inter-preted this to apply as an operating limit for the plant and consider that it introduces a conflict to the established conditions for plant operation as identified in the plant

' technical specifications. A factor of 3 is quite often used in ASME Code Design guidelines. These Code practices apply to.the design of hardware and to the analyses done on these designs. Conditions which occur during operation of the i equipment and which may affect the equipment so that design

. values no longer apply, are not directly addressed by the

initial Code requirements. That is cne reason why. plant

, technical specifications have been generated to establish

, safe ~1imits of operation for power station _ equipment. The

~; ASME Code is not applicable to the operational criteria of steam generator tubing. Our tubing design and tubing in the i design condition has margins in excess of 3. In summary, we satisfy the margin of 3 if it were used in.a Code sense as new equipment design. Moreover, we do not believe that this margin should be utilized as a limiting condition for normal operation.

$ . t A00.7.4529 A1.121-1

B/B-FSAR AMENDMENT 37 MARCH 1982 Position C.2.b In cases where sufficient inspection data exists to establish a degradation allowance, the rate used will be an average time-rate determined from the mean of the test data. The combined effect of these requirements would be to establish a maximum permissible primary-to-secondary leak rate which may be below the threshold of detection with current methods of measurement.

Westinghouse has determined the maximum acceptable length of a through-wall-crack based on secondary pipe break accident loadings which are typically twice the magnitude of normal operating pressure loads. Westinghouse will use a leak rate associated with the crack size determined on the basis of acci-dent loadings.

Position C.3.e(6)

Westinghouse will supply computer code names and references rather than the actual codes.

Position C.3.f(l)

The 40% T.S. limit is a reference limit for Westinghouse steam generators. Regulatory Guide 1.121 analyses have not been completed for model D4 and D5 steam generators used in Byron /

Braidwood. These analyses will be completed prior to first refueling and at that time the T.S. limits will be re-evaluated and this information can then be included in the Regulatory Guide 1.121 position, if necessary.

Position C.e.f(4)

Where requirements for minimum wall are markedly different l for different areas of the tube bundle, e.g., U-bend area versus i straight length in Westinghouse designs, two plugging limits may be established to address the varying requirements in a manner which will not require unnecessary plugging of tubes.

A1.121-2 A0014530

B/B-FSAR AMENDMENT 37 MARCH 1982 REGULATORY GUIDE 1.122 Revision 1, February 1978 DEVELOPMENT OF FLOOR DESIGN RESPONSE SPECTRA FOR SEISMIC DESIGN OF FLOOR-SUPPORTED EQUIPMENT OR COMPONENTS The plant design conforms to the regulatory position as described in Subsection 3.7.2.

l l

l j

A1.122-1 A0014531

1 B/B-FSAR AMENDMENT 37 MARCH 1382 l

REGULATORY GUIDE 1.123 l

Revision 1, July 1977 l QUALITY ASSURANCE REQUIREMENTS FOR CONTROL OF PROCUREMENT OF ITEMS AND SERVICES FOR NUCLEAR POWER PLANTS The Applicant complies with this regulatory guide. Refer to Chapter 17.0 of the FSAR for further information on the Commonwealth Edison Quality Assurance Program.

A9014532 A1.123-1 l

B/B FSAR AMENDMEh'r 38 MAY 1982

' REGULATORY GUIDE 1.124 Revision 1, January 1978 SERVICE LIMITS AND LOADING COMBINATIONS FOR CLASS 1 LINEAR TYPE SUPPORTS The design of the Byron /Braidwood NSSS components supports is in cun.pliance with all of the applicable regulatory positions contained in Regulatory Guide 1.124. See Sections 3.6, 3.7, 3.9, and 3.10 for further information.

l l

l f

i 1 A1.124-1 A003.4533 l

B/B-FSAR AMENDMENT 38 MAY 1982 i

REGULATORY GUIDE 1.125 Revision 1, October 1978 PHYSICAL MODELS FOR DESIGN AND OPERATION OF HYDRAULIC STRUCTURES AND SYSTEMS FOR NUCLEAR POWER PLANTS i

This regulatory guide is applicable to construction permit applications docketed after March 1977. The Byron /Braidwood construction permit application was docketed prior to this date. However, there are no safety-related hydraulic model tests used in the designs.

J 40014534 A1.125-1

B/B-FSAR AMENDMENT 37 i MARCH 19 82 REGULATORY GUIDE 1.126 Revision 0, March 1977 AN ACCEPTABLE MODEL AND RELATED STATISTICAL METHODS FOR THE ANALYSIS OF FUEL DENSIFICATION

- The Guide states clearly that, "The model presented in ...

this guide is not intended to supersede NRC approved vendor models." Westinghouse prefers to use the densification model presented in WCAP-8218 (Proprietary) which has been approved s by the NRC. WCAP-8219 (Non-Proprietary) and WCAP-8264 are

, companions to the approved versions.

)

i 1

'A0014535 i

A1.126-1 L

B/B-FSAR AMENDMENT 38 MAY 1982 l

REGULATORY GUIDE 1.127 Revision 1, March 1978 l

INSPECTION OF WATER-CONTROL STRUCTURES ASSOCIATED WITH NUCLEAR POWER PLANTS The Applicant meets the requirements of the Regulatory Position of this guide.

i l

l l

l >

l 4

f i

L 1

t t

A1.127-1 A8)O14536

i i

B/B-FSAR AMENDMENT 37 MARCH 1982 REGULATORY GUIDE 1.128 Revision 1, October 1978 INSTALLATION DESIGN AND INSTALLATION OF LARGE LEAD STORAGE BATTERIES FOR NUCLEAR POWER PIANTS The Applicant conplies with the requirements of this guide with the exceptions and/or clarifications to the Regulatory Positions identified and justified below:

Regulatory Position C-1 In Subsec tion 4.1.4, " Vent ilation , " instead of the second sentence, the following should be used:

"The ventilation system shall limit hydrogen concentra-tion to less than two percent by volume at any location within the battery area."

Applicant's Position The ventilation requirements set forth in IEEE Std. 484-1975 are adequate.

Justification of Applicant's Position IEEE 484-1975 requires that the ventilation system limit hydrogen accumulation to less than 2% of the total volume of the battery area. This Regulatory Position would require that hydrogen accumulation be limited to less than 2% at any location within the battery area. The ventilation requirements as set forth in IEEE 484-1975 are entirely adequate. The "2% at any location" requirement would be almost impossible to verify and might even require the installation of ducts, vanes, and/or auxiliary fans so as to ensure that every " nook and cranny" is thoroughly purged.

The battery area ventilation system is designed to maintain the hydrogen concentration below 2% with a "run-away" charger (i.e., a charger delivering its full-rated output into a fully-charged battery, thereby causing gasing of all cells). Thus, any significant hydrogen build-up in the A1.128-1 A003AS37

B/B-FSAR AMENDMENT 37 MARCH 1982 battery area would require two failures (a failure of the ventilation system, and a failure of the charger), both of which will be annunciated in the main control room.

Regulatory Position C-2 In Subsection 4.2.1, " Location," Item 1, the general require-ment that the battery be protected against fire should be I supplemented with the applicable recommendations in Regula-tory Guide 1.120, " Fire Protection Guidelines for Nuclear Power Plants."

Applicant's Position The reference to Regulatory Guide 1.120 is inappropriate l because this Regulatory Guide is presently only in the

" comment" stage.

Justification of Applicant's Position The battery location and protection against fire will be described in the Fire Protection Report in Response to Branch Technical Position APCSB 9.5.1 in lieu of Regulatory Guide 1.120. The location and fire protection requirements set forth in IEEE 484-1975 are adequate.

In reference to Regulatory Guide 1.120, Revision 1, (November 4

1977), Section C.6(g), Page 20, " Safety-Related Battery Rooms," Applicant's comments are as follows:

(a) This paragraph seems to imply that all safety- '

related batteries are to be located in separately-enclosed rooms. It is Applicant's position that it should not be necessary that battery rooms be separated from other areas of the plant by barriers having a minimum fire rating of three hours. Such barriers would be necessary only if the batteries were in a separate fire protection

, zone. There is nothing wrong with a design wherein l the battery is located in an open area so long as the battery is protected from mechanical damage;

e.g., the battery may be located in an Electrical Equipment Room but protected by an enclosing fence.

, (b) The location of de switchgear and inverters in the Electrical Equipment Room described above is a satisfactory arrangement.

! Regulatory Position C-3

Items 1 through 5 of Subsection 4.2.2, " Mounting," should be supplemented with the following

A1.128-2 A003A538

B/B-FSAR AMENDMENT 37 MARCH 1982 "6. Restraining channel beams and tie rods shall be electrically insulated from the cell case and shall also be in conformance with Item 2 above regarding moisture and acid resistance."

In addition, the general requirement in Item 5 to use IEEE 344-1975 should be supplemented by Regulatory Guide 1.100, l

" Seismic Qualification of Electric Equipment for Nuclear Power Plants."

Applicant's Position j

Restraining channel beams and tie rods need not be electri-cally insulated from the cell case.

Justification of Applicant's Position The expense for the addition of electrical insulation to the restraining. channel beams and tie rods is unwarranted.

Heat from an accident that can damage lead plates and vapor-ize electrolyte could also melt insulation on restraining channels and tie rods. In addition, rubber or plastic for insulation purposes will significantly increase the combustible fuel loading in the battery area and thus add to the fire hazard.

l

\

l l

i l

A1.128-3 A0014533

B/B-FSAR AMENDMENT 37 MARCH 1982 REGULATORY GUIDE 1.129 l Revision 1, February 1978 MAINTENANCE, TESTING AND REPLACEMENT OF LARGE LEAD STORAGE BATTERIES FOR NUCLEAR POWER PLANTS The Applicant complies with the Pegulatory Position as described in Subsection 16.4.8.2.3.2.

I A1.129-1 9 03/,540

B/B-FSAR AMENDME19T 37 l MARCH 1982 REGULATORY GUIDE 1.130 Revision 1, October 1978 j l

DESIGN LIMITS AND LOADING COMBINATIONS FOR CLASS 1 PLATE-AND-SHELL-TYPE COMPONENT SUPPORTS l

The design of Byron /Braidwood NSSS component supports is in compliance with all of the applicable regulatory positions contained in Regulatory Guide 1.130.. Refer to Section 3.8 for further information.

l A1.130-1 A0014541 l

l

B/B-FSAR AMENDMENT 37 MARCH 1982 REGULATORY GUIDE 1.131 1

Revision C, August 1977 QUALIFICATION TESTS OF ELECTRIC CABLES, FIELD SPLICES AND CONNECTIONS FOR LIGHT-WATER-COOLED NUCLEAR POWER REACTORS The Applicant complies with the Regulatory position with the following comments and exceptions keyed to paragraph numbers in the position:

1. Regulatory Position C-1 The position states that in lieu of Section 1.3.4.2.3 of IEEE 383, "Other Design Basis Events", the following should be used: "The remainder of the complete spectrum of design basis events (e.g. , events such as a steam line break) shall be considered in case they represent different types of more severe hazards to cable operation."

Applicant Reponse: All safety-related cable is qualified for the anticipated environments detailed in Section 3.11 of the B/B-FSAR. Steamline breaks are addressed in Subsection 3.6.1.3 of the B/B-FSAR.

I l 2. Regulatory Position C-10 The PoFition States that in lieu of the first sentence of l Section 2.5.4.4.1 of IEEE 383, the following should be used:

The ribbon gas burner shall be mounted horizontally such that the flame impinges on the specimen midway between the tray

, rungs and so that the burner face is in front of and 4 inches from the cable and approximately 2 feet above the bottom of the tray.'

l Applicant Response: The ribbon gas burner was mounted so that the burner face was in front of and 3 inches from the l cable, as set forth in IEEE 383, Section 2.5.4.4.1.

i

3. Regulatory Posit!.on C-11

, The position states that in lieu of Section 2.5.4.4.3 of I

IEEE 383 the following should be used: " Flame size will normally be achieved when the propane flow is 27.8 standard ft per hour and the air flow is 139 standard ft per hour."

l A1.131-1 'A0014542

9/B-FSAR AMENDMENT 18 JANUARY 1979 Applicant Response: Flame size was achieved using the schematic arrangement and pressures as set forth in IEEE 383, I Section 2.5.4.4.3.

1 l

i i

(

4 l

1 f

I A0014543 r

A1.131-2

B/B-FSAR AMENDMENT 38 MAY 1982 l

! - REGULATORY GUIDE 1.132 l

l Revision 1, March 1979 i

SITE INVESTIGATIONS FOR FOUNDATIONS OF NUCLEAR POWER PLANTS I. BYRON STATION 1

All Byron Station site investigations were performed prior to June 1,1978, with the exception of eleven borings (IC-1 through IC-ll) performed on June 12, 13, and 14, i

1978, three borings on December 14, 15, and 16, 1981, and four borings on March 17, 18, and 19, 1982, along a portion of the essential service water pipeline. The

. site investigations of these borings conform to the guide-lines set forth in the Regulatory Guide. However, the i

site investigations performed by the Applicant prior to the date of the Regulatory Guide implementation conform to the guidelines set forth in this Regulatory Guide in that the investigation methods conform to the ASTM (American

! Society for Testing and Materials) procedures or other

generally accepted procedures for foundation investigations.

l For details see Byron Sections 2.4 and 2.5.

II. BRAIDWOOD STATION All Braidwood Station site investigation work was performed prior to June 1, 1978. However, the site investigations

, performed by the Applicant prior to the date of the Reg-l ulatory Guide implementation ~ conform to the guidelines l set forth in this Regulatory Guide in that the investigation 7

methods conform to the ASTM (American Society for Testing I and Materials) procedures or other generally accepted l procedures for foundation investigations. For details

see Braidwood Sections 2.4 and 2.5.

l l

i l

I A1.132-1 10014644 '

1

B/B-FSAR AMENDMENT 37 MARCH 1982 Regulatory Guide 1.133 Revision 0, September 1977 LOOSE-PART DETECTION PROGRAM FOR THE PRIMARY SYSTEM OF LIGHT-WATER-COOLED REACTORS The Loose Parts Detection System is in compliance with the regulatory position with the following exceptions and clarifications keyed to paragraph numbers in the regulatory position:

Section C l.a. Sensor Location Byron /Braidwood is in compliance with this section.

Section C.l.b System Sensitivity The manufacturer states that preliminary tests on the system demonstrate compliance with this section.

Section C.l.c. Channel Separation The sensors, cables, and line drivers are all physically separated from each other, but the twisted-shielded pairs running back to the control room are routed in the same e cable division.

Section C.l.d. Data Acquistion System Byron /Braidwood is in compliance with this section.

Section C.l.e. Alert Level j Byron /Braidwood is in compliance with this section.

j Section C.l.f. Capability for Sensor Channel Operability Tests

There is no. specific channel test for the system, but sensors

, on the reactor can be checked when the rods are moved.

i Section 1.C.g. Operability for Seismic and Environmental conditions J

The Loose Parts Monitoring System has not been demonstrated i

Al.133-1 f0014646 I.

B/B-FSAR AMENDMENT 19 MARCH 1979 to be capable of performing its function following all seismic events that do not require plant shutdown, up to and including the Operating Basis Earthquake (OBE).

Section 1.C.h. Quality of System Components The components have not been demonstrated to have a 40-year design life, but the Regulatory Guide penmits setting up a replacement schedule to replace these items. The Byron /

Braidwood design permits this.

Section 1.C.i. System Repair Byron /Braidwood is in compliance with this section.

r l

l-A1.133-2

, B/B-FSAR AMENDMENT 37 MARCH 1982 REGULATORY GUIDE 1.134 Revision 1, March 1979 g r*~ MEDICAL EVALUATION OF NUCLEAR POWER PLANT PERSONNEL REQUIRING OPERATOR LICENSES The Applicant complies with the Regulatory Position of this guide.

4 i

i l

l l

i p.

I A1.134-1 A0014547 I . - - . --- , - . - .- _ __

B/B-FSAR AMENDMENT 37 MARCH 1982 1

REGULATORY GUIDE 1.135 Revision 0, September 1977 NORMAL WATER LEVEL AND DISCHARGE AT NUCLEAR POWER PLANTS The plant design conforms to the regulatory positions as de-scribed in Subsections 2.4.3 and 2.4.11.

4 J

i i

1 i

i A1.135-1 A0014548 l

o .

'I B/B-FSAR AMENDMENT 37 MARCH 1982 REGULATORY GUIDE 1.136 Revision 0, November 1977 MATERI AL FOR CONCRETE CONTAINMENTS The plant design conforms to the regulatory Position as de-scribed in Appendix B. Regulatory position C.3 is not applicable since grouted tendon systems are not used.

f I

A1.136-1 A0014549 l - - - - - - __ _ -

B/B-FSAR AMENDMENT 38 MAY 1982 REGULATORY GUIDE 1.137 Revision 1, October 1979 FUEL-OIL SYSTEMS FOR STANDBY DIESEL GENERATORS The Applicant complies with the requirements of this regulatory guide. Refer to Subsection 16.4.8.1.1.2.b. l.

I t

l l

l l

l i

i I

A1.137-1 AOMd550

B/B-FSAR AMENDMENT 38 MAY 1982 i

l REGULATORY GUIDE 1.138 1

Revision 0, April 1978

• k LABORATORY INVESTIGATIONS OF SOILS FOR ENGINEERING ANALYSIS AND DESIGN OF NUCLEAR POWER PLANTS I. BYRON STATION All of the Byron Station laboratory tests on soils and rocks for determining soil and rock properties were performed prior to the Regulatory Guide implementation date of December 1, 1978, with the exception of laboratory tests of soil along a portion of the ESW pipeline in 1982. The laboratory tests performed in 1982 conform to the requirements of Regulatory Guide 1.138. However, the Applicant's laboratory investigations of soils and rocks prior to December 1, 1978 conform to the guidelines set forth in the Regulatory Guide in that the ASTM (American Society for Testing and Materials) procedures or other generally accepted pro-cedures were used in performing the laboratory testing. .

For details, see Byron Subsections 2.5.4 and 2.5.5. '

II. BRAIDWOOD STATION i

i The only laboratory tests performed since December 1, 1978 at the Braidwood Station were on soils along a portion

, of the essential service water pipeline and several areas

! of the exterior dike embankment. The essential service i water pipeline is Safety Category I, the Exterior Dike 4

Embankment is non-safety-related.

The laboratory testing of soils by the Applicant since December 1, 1978 conforms to the requirements of Regulatory Guide 1.138. (No rock has been tested since December 1, j 1978.) The laboratory testing of soil and rock by the l Applicant prior to December 1, 1978 conforms to the guide-i lines set forth in Regulatory Guide 1.138 in that the ASTM (American Society for Testing and Materials) procedures or other generally accepted procedures for laboratory testing of soil and rock were used. For details see Braid-wood Subsections 2.5.4, 2.5.5, and 2.5.6.

Al.138-1 ,C303.4551 i . - . . .

l B/B-FSAR RMEEDMENT 37 MARCH 1982 REGULATORY GUIDE 1.139 Revision 0, May 1978 GUIDANCE FOR RESIDUAL HEAT REMOVAL The Applicant complies with the requirements of this guide, l in that the Byron and Braidwood Stations are designed for

" Safe .hutdown" concurrent with a single failure in one of the redundant ESP divisions. However, the applicant defines the term " Safe Shutdown" as meaning " Hot Standby."(1) Refer to Subsection 5.4.7 for further information, r

i i

i l

i l

l III As defined in the Technical Specification, FSAR Chapter 16.0.

l i

l A1.139-1 A0014552

B/B-FSAR AMENDMENT 42 MAY 1983 REGULATORY GUIDE 1.140 Revision 0, March 1978 DESIGN, TESTING AND MAINTENANCE CRITERIA FOR NORMAL I VENTILATION EXHAUST SYSTEM AIR FILTRATION AND ADSORPTION UNITS OF LIGHT-WATER-COOLED NUCLEAR POWER PLAN'TS The design of the non-safety-related filter systems meet the requirements of this guide, except as noted below. The excep-tions are keyed to paragraph numbers in the Regulatory Position.

la and lb - The equipment and components (excluding charcoal, filter pads and separator pads) are designed to withstand a maximum of 40-year integrated radi-ation dose and worst-case anticipated continuous service, rather than 40 years of continuous service.

2a - All of the exhaust filter systems contain prefilters, HEPA filters, fan and associated instrumentation.

Charcoal adsorbers are only used when iodine is anticipated to be present, with heaters for over 70% relative humidity air streams.

2b - The purge system exhaust filter train is designed for 43,900 cfm. Filter efficiency will be tested at this capacity. The filter train consists of two banks with a grating between the filter banks. Each filter bank is three filters high and'seven wide.

2f and 3f - Filter Housings All non-ESF filter housings, exclusive of the TSC and post-LOCA purge units are at negative pressure with respect to their surroundings, and are located in areas which are low airborne radiation environments. Any in-leakage will l not adversely affect Appendix I releases, hence, the housings will not be leak tested to the ANSI-N509-76 requirements. The TSC unit housing is located in an area where the airborne radiation level of the room air may exceed that of the air within the housing; however, it is at positive pressure with respect to the surroundings, hence, it will not be tested to ANSI-N509-76 requirements.

The post-LOCA purge unit housing will be leak tested to ANSI N509-76 requirements.

A1.140-1 A0014553

B/B-FSAR AMENDMENT 46 JANUARY 1985 2f and 3f - Ductwork All of the ductwork upstream of the non-ESF filter units is under negative pressure with respect to its surroundings. The ductwork upstream of

the TSC filter unit is located in the HVAC equip-i ment room. The quality of the equipment room environment is the same as that of the outside l .

air which is within the duct. Any in-leakage will be filtered prior to its release to the TSC environment, hence, this ductwork will not be tested to the ANSI N509-76 requirements.

, Ductwork upstream of the remaining filter units i

is located in areas of low airborne radioactivity.

i Any in-leakage will not adversely affect Appendix l I releases, hence, it will not be tested to the ANSI N509-76 requirements.

All positive pressure ductwork associated with the non-ESF filter units will not be leak tested, however, the off-gas system discharge ductwork and sections of miscellaneous ventilation system q ductwork passing through the control room boundary were tested in the field and leak rates were found to be within the limits of Section 4.12 of ANSI N509-76. These duct sections are repre-sentative of the duct construction used for all HVAC ductwork.

j Any ductwork sections where leakage would affect '

the habitability of the control room or TSC will be tested to the ANSI N509-76 requirements.

) 3a - The components of the heaters are manufactured and assembled as per Section 5.5 of ANSI N509-76, similar to the requirements of heaters in safety-related filter systems, but the traceability of the components is not established as it would  !

be in the case of safety-related heaters. Thus, l no complete qualification program is done, however, i

i all air heaters (ESF and non-ESF) are purchased using the same design, construction, and performance '

specifications.

i Sb - Airflow distribution and air-aerosol mixing tests  ;

i will not be performed on the non-entry type filter

units. Airflow distribution tests will be performed i

on all entry-type filter trains to ensure that the airflow through any individual filter element does not exceed 120% of the elements rated capacity.

A1.140-2 A001.4554

B/B-FSAR AMENDMENT 37 MARCH 1982 REGULATORY GUIDE 1.141 Revision 0, April 1978 i

CONTAINMENT ISOLATION PROVISIONS FOR FLUID SYSTEMS The Applicant complies with the requirements of this Regulatory Guide, as further explained below:

a. Phase "A" and Phase "B" Conditions are different from those listed. Refer to Figure 7.2-1. *

b. There are differences between various figures in Appendix "B" and containment isolation features for various systems. Refer to diagrams ot the various systems. Appendix "C" pertains to diagram legend and symbols, to which the Applicant conforms with minor exceptions. Appendix "D" pertains to a valvt maintenance program which the Applicant does not agree to implement. The Applicant agrees with the exceptions which the guide has taken to ANSI N271-1976.

5 i

A1.141-1 l

A0014555

i 1

B/B-FSAR AMENDMENT 37  ;

MARCH 1982 REGULATORY GUIDE 1.142 ,

I i

Revision 0, April 1978 t

SAFETY-RELATED CONCRETE STRUCTURES FOR NUCLEAR POWER PLANTS (OTHER THAN REACTOR VESSELS AND CONTAINMENTS)

The compliance with the requirements of this Regulatory Guide is presented in the response to Question 130.21.

i f

4 i

i l

t i

l i,

I Al.142-1 l A0014S56 l

l l

5_ 1

, i B/B-FSAR AMENDMENT 37 MARCH 1982 t

REGULATORY GUIDE 1.143 Revision 0, July 1978 DESIGN GUIDANCE FOR RADIOACTIVE WASTE MANAGEMENT SYSTEMS, STRUCTURES AND COMPONENTS INSTALLED IN '

LIGHT-WATER-COOLED NUCLEAR POWER PLANTS 1

The Applicant complies with the requirements of this guide.

j Further information is provided in Subsection 11.2.1.11.

I i

^1*1"3-1 l

As)e3AS57

-, . - - , - --._ .n_ - - - . - . . - , - . - - , - . - . .

.,.,n.- -~, . . - .- . ,

. . -.. . _ .- . . . - _ . . . _ _ _ _ . . _ _ . _._. _ . _ . _ _ _ _ ~ . . _ _ _

, i l B/B-FSAR AMENDMENT 37

MARCH 1982 j REGULATORY GUIDE 1.144 Revision 0, January 1979 AUDITING OF QUALITY ASSURANCE PROGRAMS

, FOR NUCLEAR POWER PLANTS 4 The Applicant complies with the Regulatory Position in this l guide. Refer to Chapter 17.0 for further references.

I 1

I  !

5 d

i

)

i.

I t 4

i i

i I

r

, I i

A0014S58  :

.A1.144-1

a .

] B/B-FSAR AMENDMENT 37 MARCH 1982 4-REGULATORY GUIDE 1.146 i

i i

t i Revision 0, August 1980

) OUALIFICATION OF QUALITY ASSURANCE PROGRAM i AUDIT PERSONNEL FOR NUCLEAR POWER PLANTS 1

< The Applicant complies with the Regulatory Position in this j guide. Refer to FSAR Chapter 17.0 for further references.

I t

r

'I 1 i 1

i i

j i .

l i

1 e 4

4 i

.I

! t

[

i i

A1.146-1 A0014S59 l

.c..-. - - - - - . - . - . - . -

o e i B/B-FSAR AMENDt1ENT 37

~

MARCH 1982 REGULATORY GUIDE 8.2 I Revision 0, February 2, 1973 GUIDE FOR ADMINISTRATIVE PRACTICES IN RADIATION MONITORING Administrative procedares and practices of radiation monitoring

are based on 10 CFR 20 and Regulatory Guide 8.2, Revision -

February 2, 1973.

)

I i .

i i

AOP34Sf;0 A8.2-1 l

D/O-FSAR AMENDMENT 37 MARCH 1982 I

REGULATORY GUIDE 8.3 2

i Revision 0, February 2, 1973 ,

I FILM BADGE PERFORMANCE CRITERIA Film badge performance is verified by the film badge vendor by adhering to Regulatory Guide 8.3, Revision - February 2, 1973, for film badge performance criteria.

1 1

l 2

,i 4 04d.4'561 A8.3-1 l

e >

D/B-FSAR AMENDMENT 37 MARCH 1982 REGULATORY GUIDE 8.7 Revision 0, May 1973 OCCUPATIONAL RADIATION EXPOSURE RECORDS SYSTEM occupational radiation exposure record system is based on Regulatory Guide 8.7, Revision - May 1973.

i i

t l

A8.7-1 f0014!iG2

, i AMENDMENT 43 B/B-FSAR SEPTEMBER 1983 REGULATORY GUIDE 8.8 Revision 3, June 1978 INFORMATION RELEVANT TO ENSURING THAT OCCUPATIONAL RADIATION EXPOSURES AT NUCLEAR POWER STATIONS WILL BE AS LOW AS IS REASONABLY ACHIEVABLE Regulatory Guide 8.8 describes information that is relevant to meeting the criterion that exposures of station personnel to radiation during routine operations of the station will be as low is reasonably achievable (ALARA) .

Maintaining occupational radiation doses ALARA is a function of the Health Physics Program (12.5), station design (12.3), and administrative policies (12.1.1.3). The Applicant has also used operating experience during the design phases and is utilizing supervisory personnel who have several years of operating expe: '.ence working in its licensed stations.

The Health Physics Program includes The the administrative radiation protection policies program, training, and instruction.

maintain occupational exposure ALARA and establish station organization, responsibilities, and procedures. The station design includes access control, shielding, facility design, equipment design, airborne control, crud coatrol, radiation monitoring, waste treatment, and modifications (based on operating experience).

The guidance provided by Regulatory Guide 8.8 (all issues) and by WCAP-8872, " Design, Inspection, Operation, and Maintenance Aspects of the Westinghouse NSSS to Maintain Occupational Radiation Exposures as Low As Reasonably Achievable," has been used as an aid for the radiation protection design. The Applicant believes that it complies with the guidance set forth in this regulatory guide.

A8.8-1

, 2 B/B-FSAR AMENDI1ENT 37 MARCH 1982 REGULATORY GUIDE 8.9 Revision 0, September 1973 ACCEPTABLE CONCEPTS, MODELS, EQUATIONS, AND ASSUMPTIONS FOR A BIOASSAY PROGRAM The bioassay program will be in compliance with Regulatory Guide 8.9, " Acceptable Concepts, Models, Equations, and Assumptions for a Bioassay Program". Bioassay services will be performed by either a contracted vendor or by Commonwealta

Edison personnel. When a vendor is contracted to perform bioassay, a requirement of the contract will be compliance with the approrpriate regulatory requirements. In the case of Commonwealth performing bioassay services, the bioassay program will be in compliance with the applicable regulatory requirements.

I l

1 j

J I

j A8.9-1 A ^ l

, 2 B/B-FSAR AMENDMENT 37 MARCH 1982 l l

REGULATORY GUIDE 8.10 Revision 1, September 1975 4

OPERATING PHILOSOPHY FOR MAINTAINING j OCCUPATIONAL RADIATION EXPOSURES AS LOW AS IS REASONABLY ACHIEVABLE The operating philosophy for maintaining occupational exposures ALARA is based on Regulatory Guide 8.10 to the degree considered reasonable by the respective stations.

i

}

i i

)

i i

Age 14Sr5 A8.10-1

o f B/B-FSAR A!!ENDMENT 37 MARCH 1982 4

REGULATORY GUIDE 8.12 0

Revision 0, December 1974 i

4 CRITICALITY ACCIDENT ALARM SYSTEMS Area monitors are provided near the spent fuel storage pool to alarm locally and in the main control room. These monitors will respond to radiation in the event of a criticality accident in the new fuel storage area. These monitors meet the intent of Regulatory Guide 8.12 and are further described in Subsection 11.5.2.2.6.

i d

A8.12-1

,00].45cc A

o .s i

B/B-FSAR AMENDMENT 37 MARCl! 1982 i

REGULATORY GUIDE 8.15 Revision 0, October 1976 ACCEPTABLE PROGRAMS FOR RESPIRATORY PROTECTION The Respiratory Protection Program is organized and imple-mented in accordance with the provisions of 10 CFR 20.103 and Regulatory Guide 8.15.

b l

l

, A8.15-1 'AI)(').4{5b7 l

i

fo 1 B/B-FSAR AMENDMENT 37 tiARCH 1982 REGULATORY GUIDE 8.19 Revision 0, May 1978 OCCUPATIONAL RADIATION DOSE ASSESSMENT IN LIGHT-WATER REACTOR POWER PLANTS DESIGN STAGE MAN-REM ESTIMATES The dose assessment objectives of Regulatory Guide 8.19 have been included in the FSAR as indicated below:

Item C.(1)

The occupational radiation exposure estimates are in Section 12.4.4.

Item C.(2)

B/B's radiation exposure assessment bases are described in Subsection 12.1.2.3, Subsection 12.5.3, Q331.2, Q331.5, Q331.7, and 0331.9.

Item C.(3)

Design changes which have resulted from Commonwealth Edison's dose assessment process are included in Subsection 12.1.2.3 and Q331.3.

i l

A8.19-1 Q A0014568

B/B-FSAR AMENDMENT 19 MARCH 1979 APP ENDIX B - CONSTRUCTION MATERI AL STANDARDS AND OUALITY CONTROL PROCEDURES TABLE OF CONTENTS PAGE B.0 CONSTRUCTION M ATERI AL STANDARDS AND OUALITY CONTROL PROCEDURES B.1-1 B.1 CONC?ETE STANDARDS B.1-1 B.1.1 Gene ral B.1- 1 B. I. 2 Material Requirements and Quality Control B.1-1 B.1.2.1 Ce me nt B.1-1 E.1.2.2 Aggregates B.1- 2 B.1. 2. 3 Heavy Weight Aggregate B.1-4 B.I.2.4 Fly Ash B.1-4 9.1.2.5 Admixtures B.1- 5 B. I. 2. 6 Water and Ice B.1-6 B .1. 3 Concrete Properties and Mix Design B.1-7 l B.I.3.1 Trial Mixtures B .1- 7 B. I. 3. 2 Design Mixtures B.1-7 B.1.3.3 Adiustnent of Design Mixtures B.1-8 B.1.3.4 Groet B.1- 8 B. I . 3. 5 Additional Concrete Testing for Concrete Used in Containment B.1- 8 E.1.3.6 Heavyweight Concrete B.1-9 B.1.4 Formwork B.1-9 B.1. 5 Joints and Embedded Items B.1- 9  ! l B.1. 6 Bar Placement B.1-10 B.1.7 Bending or Straightening of Bars Partially Embedded in Set Concrete B.1- 11 B .1. 8 Batching, Mixing, Delivery, and Placement B.1-12 B.I.9 Witness and Inspections B.1- 12 B.1.10 Concrete Placement B.1-12 l B.1.11 Concrete Control Tests B.1-15 B.1.12 Evaluation and Acceptance of Fresh Concrete B.1- 15 B.1.13 Evaluation and Acceptance of Concrete Compression Results B.1- 16 B. I.14 Consolidation of Concrete B.1- 16 l B.1.15 Concrete Finishes B.1-17 D.1.16 Curing and Protection B.1-18 B.I.17 Preplaced Aggregate Concrete B.1- 18 B. I.18 Evaluation and Acceptance o f Concrete B.1-19 l B.2 PEINF0FCING STEEL B . 2- 1 B.2.1 Requirements for Category I Materials B.2-1 B.2.2 Reinforcing Bar Fabrication B. 2- 3 B.2.3 Cadweld Splicing B.2-3 B.2.3.1 Qualification of Operators B.2-3 l

B.0-1 gg A.'Jpygg g

1 B/B-FSAR AMENDMENT 19 MARCH 1979 TABLE OF CONTENTS (Conted)

PAGE B.2.3.2 Procedure Specifications B . 2- 3 B.2.3.3 visus 1 Examination B.2-4 B . 2. 3. 4 Sampling and Tensile Testing B.2-6

b. 3 POST-TENSIONING TENDONS B.3-1 B.3.1 General B. 3- 1 B.3.2 Materials B.3-1 B.3.2.1 Tendon Material B.3-1 B.3.2.2 Buttonheads B . 3- 1 B.3.2.3 Tendon Sheathing B.3-1 B.3.2.4 Permanent Corrosion Protection B.3-1 B.3.2.5 Anchor Heads B . 3- 2 I B.3.2.6 Pearing Plates and Shims B.3-2 B.3.3 Quality Control B . 3- 2 B. 3. 3.1 Test ing B.3-2 B.3.3.1.1 Tendon Tests B.3-2 B.3.3.1.2 Tests on Wires and Buttonheads B . 3- 2 B.3.3.1.3 Tests on Corrosion Preventative Grease B.3-3 l B.3.3.1.4 Anchorage Hardware Tests and Inspection B.3-3 B.3.3.2 Fabrication Tolerances B . 3- 3 B. 3. 3. 3 Field Installation Tolerances B.3-3 B.3.3.4 Corrosion Protection B . 3- 4 l B.4 STPUCTUPAL STEEL B. 4- 1 B. 4.1 Structural Steel Materials B.4-1 B.4.2 Structural Steel Connections and Connection Material B. 4- 1 B. 4. 2.1 Bolted Connections B.4-1 B.4.2.2 Welded Connections B . 4- 1 B. 4. 3 Quality Control B.4-1 B.4.3.1 Gene ral B.4-1 B.4.3.2 Testing and Inspection of Weldments B. 4- 2 B. 4. 3. 3 Fabrication B.4-2 B.5 CO'7TAINMENT LINEF WITHIN THE CONTAINMENT BACKED BY CONCRETE B.5-1 B.S.1 Ge ne ral B . 5- 1 B. 5. 2 Materials B.5-1 B.S.3 Quality control B.5-1 B.5.3.1 Testing of Welds B . 5- 1 B.S.3.1.1 General B.5-1 B.S.3.1.2 Liner Plate Seam Welds B.5-1 B.S.3.1.2.1 Fadiographic Examinations B . 5- 1 B.5.3.1.2.2 Ultrasonic Examinations B.5-2 B.S.3.1.2.3 Magnetic Particle Examination B.5-2 B.0-li igy-l()

t \

B/B-FSAR AMENDMENT 19 MARCH 1979 TABLE OF CONTENTS (Cont'd) 2AGl B.S.3.1.2.4 Liquid Penetrant Examination B . 5- 2 B.S.3.1.2.5 Vacuum Box Soap Bubble Test B.5-2 B.S.3.1.3 Leak Test channels B.5-2 B.5.3.2 Fabrication and Installation B . 5- 2 B.S.3.2.1 General B.5-2 B.5.3.2.2 Welding Qualification B. 5- 3 B.6 COWTAINMENT STEEL BOUNDARY NOT BACKED BY CONCRETE B.6-1 B.6.1 Materials B. 6- 1 B.6.2 Quality control B.6-1 B.6.2.1 Testing B.6-1 B.6.2.1.1 General B . 6- 1 B.6.2.1.2 Testing of Welds B. 6-1 B.6.2.2 Fabrication and Installation B. 6- 2 B.6.2.2.1 General B.6-2 B.6.2.2.2 Qualification of Welders B.6-2 B.7 STAINLESS STEEL POOL LINERS B.7-1 B.7.1 flaterials B. 7- 1 b.7.2 Welding B.7-1 B.7.3 Erection Tolerances B.7-1 B.9 CfrHER STAINLESS STEEL ELEMENTS B.8-1 B.9 NUCLEAP STEAM SUPPLY SYSTEM (NSSS) COMPONENT SUPPOPT STEEL B.9-1 B . 9.1 General B . 9- 1 B.9.2 Steel Materials B.9-1 B.9.3 Welding Qualifications B.9-1 B.9.4 Quality control B . 9- 1 B.9.4.1 General B.9-1 B.9.4.2 Lamination Tests B.9-1 B.9.4.3 Nondestructive Examination of Welds B . 9- 1 B.9.5 Fabrication and Installation B.9-1 B . 9. 5.1 Installation Tolerances B.9-2 B.0-lii N '

t 'N l

B/8-FSAR AMENDMENT 19 i MARCH 1979 l

APPENDIX B - CONSTRUCTION MATERIAL STANDAPDS AND QUALITY CONTROL PROCEDURES LIST OF TABLES NUMBER _ TITLE Z8EX B .1- 1 A'.r Content B.1- 20 B.1-2 Limits for Slump B.1-21 B.1-3 Placing Temperature B.1-22 B.1-4 Concrete Compression Testing B.1- 2 4 B.1-5 Concrete Testing B.1-25 B.1-6 Gradation of Heavyweight Aggregate B.1- 26  !

B.9-1 Material for NSSS Component Supports B.9-3 B.0-iv

t T B/B-FSAR AMENDMENT 42 MAY 1982 s

APPENDIX B CONSTRUCTION MATERIAL STANDARDS AND QUALITY  !

CONTROL PROCEDURES l B. O CONSTRUCTION MATERIAL STANDARDS AND QUALITY CONTROL PROCEDURES B.1 CONCRETE STANDARDS B. I.1 General All concrete work done conformed to the requirements as cited from the codes, Standards and Recommended Practices as listed in Table 3.8-2 with the exceptions and additional requirements indicated in this Section B.1.

B.1.2 Material Requirements and Quality control B.1. 2.1 Ce ment Portland Cement, Type II, was used and cc4 forms to all applicable requirements of " Specification for Portland Cement" C150. Portland Cement, Type I, conformed to all the(ASTM standard  !

chemical requirements and standard physical requirements listed in Tables 1 and 2 respectively of ASTM C150. Type I cement was used on a limited basis in Category I structures other than the containment.

Qualification Tests preliminary to mix design were performed on every source of cement for conformance with ASTM C150.

The cement supplier furnished certification with each shipment of cement to the project site for the following ASTM tests:

a. ASTM C114, " Chemical Analysis of Rydraulic cement,"

including actual Na 20 content and requirements for tricalcium silicate and tricalcium aluminate as specified in Table 1A of ASTM C150.

b. ASTM C109, " Test for Compressive Strength of Hydraulic Cement Mortars" (results were forwarded within 30 days after delivery) ,

c. ASTM C204, " Test for Fineness of Portlant Cement by Air Permeability Apparatus," and

d. ASTM C266, " Tests for Time of Setting of Hydraulic Cement by Gilmore Needles," or C191, " Tests for Time of Setting of Hydraulic Cement by Vicat Needle."

D.1-1 AUU2457:3

s-t A l B/B-FSAR RMENDMENT 42 1

MAY 1983 Control Testing was performed for the following ASTM tests, based on a frequency of every 1200 tons:

a. ASTM C114,

b. ASTM C266 or C191,

c. ASTM C151, "Tes t for Autoclave Expansion of Portland i Cement," and

d. ASTM C204.

All cement was stored in accordance with the applicable requirements of Section 2.5.1 of ACI 301, " Specification for l Structural Concrete for Buildings."

B.1.2.2 Aqqrega tes Fine and coarse aggregates conformed to " Standard Specification for Concrete Aggregates" (ASTM C33) and to the following. l

a. Size Numbers 57 or 67 were used (ASTM C33). l

b. Coarse aggregate contained less than 155 (by weight) flat and elongated particles as determined by l CRD-C119 , " Method of Test for Flat and Elongated i Particles in Coarse Aggregate."

Samples of aggregate were obtained in accordance with ASTM D75,

" Sampling Aggregates," and the following Qualification Tests preliminary to mix design were performed on each source and type of aggregate proposed for use:

a. ASTM C136, " Sieve or Screen Analysis of Fine and Coarse Aggregates,"

b. ASTM C117, " Materials Finer Than No. 200 Sieve in Mineral Aggregates by Washing,"

c. ASTM C40, " Organic Impurities in Sands for Concrete,"

d. ASTM C87, "Effect of Organic Impurities in Fine Aggregate on Strength of Mortar,"

e. ASTM C88, " Soundness of Aggregates by Use of Sodium l Sulfate or Magnesium Sulfate,"

f. ASTM C142, " Clay Lumps and Friable Particles in

• Aggrega te s,"

g. ASTM C123, " Lightweight Pieces in Aggregate,"

B.1-2 AOUFMM

m t N B/E-FSAR AMENDMENT 19 MARCH 1979

h. ASTM C131, " Resistance to Abrasion of Small Size l Coarse Aggregate by Use of the Los Angeles Machine,"

i. ASTM C235, " Scratch Hardness of Coarse Aggregate  !

Particles,"

j. ASTM C127, " Specific Gravity and Absorption of Coarse l Aggregate,"

k. ASTM C128, " Specific Gravity and Absorption of Fine l Aggrega te ,"

1. ASTM C29, " Unit Weight of Aggregate," l

m. ASTM D1411, " Water-Soluble chlorides Present as l Admixes in Graded Aggregate Road Mixes,"

n. ASTM C289, " Potential Reactivity of Aggregates  !

(Chemical Method) ,"

o. ASTM C295, " Petrographic Examination of Aggregates l for concrete," and

p. CRD-C119.

The following control tests were performed during periods of casting of concrete to ascertain conformance with ASTM C33,

" Specifications for Concrete Aggregates," at the frequencies indicated:

a. ASTM C117 and C29 - daily;

b. ASTM C136 and C40 (C87 if C40 f ailed) - daily; l

c. ASTM C566, " Total Moisture Content of Aggregate by j Drying" - twice daily;

d. and CRD-C119, C142, C123, C235, C127 and C128 are l performed monthly during production; and

e. ASTM C131 or C535, 'Pesistance to Abrasion of Large l Size coarse Ar,gregate by Use of the Los Angeles Machine," C289 and C88 every 6 months.

If on aggregate sample failed any of these tests, two additional samples were taken immediately and the test for which the original sample did not meet specification requirements was repeated on each. If both samples met requirements, the material was accepted. If one or both of the retests failed, production was halted and Sargent s lundy was notified, and determined the

! necessary action required.

i B.1-3 AMEN

v B/B-FSAR AMENDMENT 42 MAY 1981 l

Samples for tests were in accordance with ASTM 075, Paragraph 3.3.3, with the following modification: the gradation tests for each source and type of aggregate proposed for use that day were i performed on samples collected and blended into one combined sample from four locations in that portion of the stockpile intended for use that day.

Control, handling and storage of aggregates, were in accordance with Section 2. 5. 2 of ACI 301.

B.I.2.3 Heavy Weicht AQQreQate Heavy weight aggregate conformed to ASTM C637, " Aggregates for  !

Radiation-Shielding Concrete," Sections 5 and 6, except that at the Byron Station a larger peracntage of material finer than sieves No. 100 and No. 200 was allowed, based on Section 4.1 of l ASTM C637 and actual placing tests. Heavy weight aggregates I also conform to the following.

Test Frequency: One complete set of qualification tests of each type of aggregate and one complete test at the time aggregate was delivered at the project site were performed. Sampling was performed in accordance with Paragraph 7.1 of ASTM C637.

B.1.2.4 Fly Ash Fly ash conformed to the Class C and F mandatory requirements of ASTM C618, " Standard Specifications for Fly Ash and Raw or Calcined Natural Pozzolan for Use as a Mineral Admixture in l Portland Cement Concrete," except that in Table 1, " Chemical Requirements," the loss on ignition was & maximum of 65.

Qualification tests preliminary to mix design were performed on every source of fly ash for conformance with the above specification.

i Certified Material Test Reports were furnished by the fly ash supplier who certified that delivered fly ash had been tested and did meet the mandatory requirements of ASTM C618, as modified l above. This certification included the test results and was supplied with delivery of fly ash for mix designs and with each 2000 tons delivered to the project.

Control Testing was performed to ascertain conformance with ASTM t i

' C618 for the following ASTM tests, based on a frequency of every 200 tons: -

a. ASTM C114 - Loss on ignition and Sulfur Trioxide

b. ASTM C430 - Amount retained on No. 325 selve.

Storage of Fly Ash was as specified in Section 2.5.1 of ACI 301 B.1-4 A00]A37c

B/B-FSAR AMENDMENT 42 MAY 1983 B.1. 2. 5 Ad mixtures Air-entraining admixtures conformed to " Specification for Air-Entraining Admixture for Concrete" (ASTM C260), including

" Optional Uniformity Pequirements" in Section 5. Air-entraining admixtures containing more than 1% chloride ions were not used.

The air entrained admixture supplier furnished certified Material

, Test Reports which state that the admixture was tested in accordance with ASTM C260 and satisfied both of these above l additional requirements. This certification includeo the manufacturer's statements as described in Sectionc 4.1, 4. 2, and 4.3 cf ASTM C260, and also included the results of the following tests performed on a composite sample from each shipment:

a. Infrared spectrophotometry

b. pH value

c. Solid content

d. Chloride ion content

e. Specific Gravity.

Chemical admixtures conformed to " Specification for Chemical Admixtures for Concrete" (ASTM C494 ) . Type A, water-reducing l

admixtures were permitted, subject to the following requirements:

a. The material was either a hydroxylated carboxylic acid base or a modified salt thereof, or a hydroxylated polymer base.

b. The material was not prepared by the addition of any chloride ions. The supplier cestified that the admixture did not contain from all sources more than 1%, by weight, of chloride ions. ,

c. The supplier furnished certified test results of specific gravity, viscosity, inf rared spectrophotometry, pH value and solids content of the material used for the project, establishing the equivalence of materials from the different lots or different portions of the same lot in accordance with Article 4.4 of ASTM C494. l Storage of Admixtures was as specified in Section 2.5.5 of ACI 301 B.1-5 AOUM577

B/B-FSAR AMENDMENT 42 MAY 1983 B. I . 2. 6 Water and Ice Mixing water and ice were clean and maximum content of chloride ion in mixing water did not exceed 500 ppm.

Qua'lification Tests preliminary to mix design were performed to ensure compliance with the requirements specified.

Control Testing was performed using the following ASTM tests:

a. ASTM D512, " Chloride Ion in Industrial Water and Industrial Waste Water" - monthly, and

b. ASTM C109, " Compressive Strength of Hydraulic Cement Mortars," C191 and C151 - every 3 months.

B.1. 3 Concrete Properties and Mix Design Concrete Mix Design conformed to ACI 211.1, " Recommended Practice for Selecting Proportions for Normal Weight Concrete,"

and to ACI 304 Title No. 68-33, " Placing Concrete by Pumping Methods," including Chapter 9 of ACI 304 l Mix Properties:

a. Slump - Concrete was proportioned to have a slump of 3 inc he s i 1 inch at 708 F as determined by ASTM C143, " Slump of Portland Cement Concrete."

b. Air content - Air content conformed to the following requirements as determined by ASTM C231, " Air Content of Freshly Mixed Concrete by the Pressure Mixture:"

Nominal Maximum size Total air content, of accrecate, in. 5, by volume 3/4 6i 1 1 5i1

c. Specified compressive strength: Structural concrete strengths and one fill concrete strength were furnished as follows:

1. 5500 psi at 91 days

2. 3500 psi at 91 days

3. 2000 psi at 28 days (fill concrete).

Fly ash content when used equaled 20% of the weight of cement.

i B .1-6

! A 00.1.-%78

~-

t N B/B-FSAR AMENDMENT 19 MAFCH 1979 B.1. 3.1 Trial Mixtures Trial mixtures having proportions and consistencies suitable for the work were made using at least three different water-cement ratios which produced a range of strengths encompassing those required for the work. All materials including the water were those used at the project site.

For each water-cement ratio, at least three compression test cylinders for each test age were made and cured in accordance with " Method of Making and curing Concrete Compression and Flexure Test Specimens in the Laboratory" (ASTM C192) . They were tested for strength at 7, 28, and 91 days , in accordance with

" Method of Test for Compressive Strength of Cylindrical Concrete Specimens" (ASTM C33; . From the results of these tests, curves were plotted showing the relationship between water-cement ratios and compressive strength.

B.1. 3. 2 Design Mixtures Until the standard deviation was calculated for each of the mixtures used, the required average strength was determined by adding 1200 psi to the required compressive strengths of 5500 psi and 3500 psi at 91 days.

This required average strcagth was entered into water-cement ratio strength curves to determine the maximum water-cement ratio.

This water-cement ratio was used with the water requirement reported from trial mixtures for the aggregate size to calculate the minimum cement content.

Adjustments in absolute volume to maintain yield were made by adjusting aggregate amounts while maintaining the sand percentage of the original trial mixture.

B.1. 3. 3 Adiustment of Design Mixtures After the accumulation of no less than 30 tests at 91 days of a mix design, these tests were evaluated by statistical methods in ACI 214 and the standard deviation was calculated. A new required average strength, fave req. wa s computed, using the higher of the values computed below:

f ave req. = fg + 1.343 o f

ave req. = fy - 500 + 2.326 o where:

fy = specified compressive strength B.1-7 f00 M 7/9 l

e s B/B-FSAR M1ENDMENT 42 MAY 1983 o = standard deviation.

With this new required average strength, the design mixtures procedure was repeated to obtain revised mix proportions using the curve for the water-cement ratio and compressive strength.

If, during the course of construction, statistical surveillance revealed that the required average was not achieved, an l investigation was performed by Sargent & Lundy to investigate the cause and determine what corrective action was necessary.

B.1.3.4 Grout Grout of proportions similar to the mortar in concrete was determined as follows:

a. A trial mix was calculated. Quantities of fly ash, fine aggregate and admixtures were computed in the same ratio to cement as in the concrete.

b. The trial mixture was performed, and the quantity of water and air entraining admixture required was determined. If water required caused the water, cement ratio of the concrete to be exceeded, cement was added until the original water-cement ratio was restored. Compressive strength of the grout was tested in accordance with ASTM C109.

B.1.3.5 Additional Concrete Testina for Concrete Used in Containment After the approval of the concrete mixtures, the following tests l

were performed:

a. Compressive strerigth at 7, 28, and 91 days (ASTM C39).

b. Static modulus of elasticity at 91 days (ASTM C469,

" Static Modules of Elasticity and Poisson's Ratio of concrete in compression") .

c. Poisson's ratio at 91 days (ASTM C469) .

d. Specific gravity at 91 days (ASTM C642, " Specific Gravity, Absorbtion, and Voids in Hardened concrete") .

e. Coefficient of thermal conductivity at 91 days (CRD-C 4 4, " Calculation of Thermal Conductivity of cement") .

l 8 1-8 AOP145SO

e s B/B-FSAp AMENDMENT 42 MAY 1983

f. Coefficient of thermal expansion at 91 days (CR D- 3 9,

" Coefficient of Linear Thermal Expansion of Concrete") .

g. Specific heat at 91 days ( ASTM C 351) .

h. Shrinkage strains up to at least 180 days after 7 and 28 days of continuous moist curing, and then drying at 70 1 20 F and 50% relative humidity ( ASTM C157) .

i. Creep strain of concrete in compression loaded after 28 and 91 days of continuous moist curing, to a sustained stress of approximately 2500 psi on test specimens drying at 70 1 20 F and 50% relative humidity and on sealed tests specimens ( ASTM C512) .

B.1.3.6 Heavyweight Concrete Heavyweight concrete mix conformed to Appendix 4 of ACI 211.1,

" Recommended Practice for Selecting Proportions for Normal Weight concrete."

B.1.4 Formwork All Formwork conformed to Chapter 4 of ACI 301 and =s l hereinafter specified.

Forms for all exposed surfaces conformed to Section 10.2.2, "Snooth Form Finish," of ACI 301. l

" Exposed surfaces" as used, meant all formed concrete surfaces exposed to view on completion of work.

All exposed projecting corners of concrete work such as piers,

, columns, eqdpment foundations, switchyard foundations, and turbine foundations were beveled.

I For exposed surfaces and exposed vertical corners of structures in contact with the ground, the smooth form finish and the vertical bevels were extended 1 foot 0 inch below finish grade.

B. I. 5 Joints and Embedded Items Joints and embedded items conformed to Chapter 6 of ACI 301 l

, including the following:

) a. For bonding methods in Sections 6.1.4.1 and 6.1.4.2

) of ACI 301, specific approval by the Purchaser or its representative was required. l

b. Horizontal construction joints in containment walls

, are cleaned by cutting the concrete surface layer and exposing the aggregate without undercutting.

1 B.1-9 10014581

B/B-FSAR AMENDMENT 42 MAY 1983 Construction joints in Category I structures were grouted immediately before placement of concrete in accordance with provisions of Section 8.5.3, ACI 301, except that no grout is required on vertical l surfaces of walls. Where keys were used, grouting of horizontal joints were not required.

c. Unformed construction joints were protected against loss of water required for curing by application, immediately after completion of construction by one of the following methods:

1. Application of damp sand or moistened fabrics kept continuously moist until placement of concrete was recommended. Prior to resumption of placement, the curing materials were completely removed from the concrete surface, in accordance with provisions of Section 8.1 of ACI 301

2. Application of curing compound containing nonfugitive pigments. Pra.or to resumption of placement, this surface was completely cleaned by sand blasting, chipping, or jack hammering until no trace of pigment remained.

B.1.6 Bar Placement sar placement conformed to the design drawings and to the applicable requirements of Section 7.2 and 7.3 of ACI 318 ; to l Chapter 8, " Placing Reinforcement Bars" of CRSI " Manual of Standard Practice," to Subsubarticles CC-4340, CC-4350, and l, CC-4360 in Section III, Division 2 of the ASME Boiler and Pressure Vessel Code, and to the following:

Ir. lieu of Section 7.3.2.1 of ACI 318 and l Paragraph 7, Chapter 8 of the CRSI Manual of Standard Practice, the following applied:

a. Clear distance to formed surfaces: For No. 3 through No. 11 bars: i 1/4 inch for straight bars, i 1/2 inch for bent bars. -

b. For No. 14 and No, 18 bars: i 1/2 inch for straight bars, i 1 inch for bent bars.

c. The cover was not reduced by more than one-third of the specified cover, nor to less than 1-1/2 inch for No. 14 and No. 18 bars at interior surfaces.

d. Spacing tolerances between parallel bara: For No. 3 through No. 18 barst' + 2 inches.

81-1 A0014SS2

, s B/B-FSAR AMENDMENT 19 MARCH 1979 B. I . 7 Bending or Straightening of Bars Partially Embedded in Set Concrete Bending or straightening of bars partially embedded in set concrete was not permitted except in isolated cases where corrective action or a field change was required and specifically approved by Sargent S Lundy.

The bend diameter conformed to the requirements listed below.

The beginning of the bend was not closer to the existing concrete surface than the minimum diameter of bend.

Bar No. 3 through No. 5 were cold bent once. Preheating was required for subsequent straightening or bending. Bars No. 6and larger were preheated for any bending.

MINIMUM DIAMETER OF BEND Bar Size Minimum Diameter of Bend No. 3 through No. 8 6 bar diameters No. 9, No. 10, No. 11 8 bar diameters No. 14, No. 18 10 bar diameters When required, preheating prior to bending or straightening was performed in accordance with the following:

a. Preheating was applied by methods which do not harm the bar material or cause damage to the concrete.

b. The preheat was applied to a length of bar at least equal to five bar diameters in each direction from the center of the portion to be bent or straightened, except that preheat was not extended below the surface of concrete. To avoid splitting the concrete, the temperature of the bar at the concrete interface did not exceed 5000 F.

c. The preheat temperature was 11000 F to 12000 F.

d.

The preheat temperature was maintained until bending or straightening was completed.

e. The preheat temperature was measured by temperature measurement crayons or contact pyrometer.

f. Precautions were taken to avoid rapid cooling of preheated bars.

for cooling.

Water was never allowed to be used l

B .1- 1 1 AW1M3 ,

t

e s B/B-FSAF AMENDMENT 42 MAY 1983 All bent and straightened bars were visually examined for cracks.

Visual examination of preheated bars was performed after the bars reached ambient temperature.

Bars straightened af ter a first bend were checked for cracking using liquid penetrant method with the following sequence:

a. The first three straightened bars.

b. One out of the next and each subsequent unit of ten straightened bars.

c. All bars with subsequent straightening or bending and all bars that failed visual examination were checked using liquid penetrant method.

d. Concrete surface was visually examined for any damage due to the preheating, bending or straightening operations.

Bars exhibiting transverse cracks were properly replaced.

B.1. 8 Ba tching, Mi xing , Delivery, and Placement Batching, mixing, and delivery equipment, including their operation, conformed to the requirements of ASTM C94, Articles l 7, 8, and 9. To the extent applicable the placement complied with the requirements of the criteria for concrete placement in Category I Structures.

B.1. 9 Witness and Inspections Prior to production, the testing agency inspected batch plant, stationary and truck mixers to verify conformance with B.l.8 above. After the concrete batch plant and mixers were placed in production, the testing agency inspected the production i facilities to verify that concrete was produced in accordance

with Section 7.2 of ACI 301.

i B.1.10 concrete Placement Air content: The allowable limits of air content was as indicated in Table B.1-1.

Slump: Concrete slump was within the allowable limits as indicated in Table B.t-2.

1 B.1-12 A00145S4

B/B-FSAR AMENDMENT 42 MAY 1983 Concrete Placing Temperature: Concrete placing temperature conformed to the criteria given in Table B.1-3.

Concrete placement conformed to the applicable requirements of Sections 8.1, 8.2, and 8. 3 of ACI 301 and Chapter 6 of ACI 304 and the following:

a. All concrete was placed in a continuous and uninterrupted operation in such manner as to form a monolithic structure, the component parts of which were integrally bonded together. No concrete was deposited which had been segregated, contaminated by foreign materials, or considered nonplastic.

b. Concrete was considered plastic if either of the following requirements were met:

1. If immediately before recommencing concrete placement a vibrator spud suspended vertically was applied to the concrete surface and it penetrated at least 6 inches into the concrete during 15 seconds of application, the concrete was considered plastic, if not, the concrete was considered nonplastic.

2. If the terperature of the concrete in place and the time interval between placement of successive batches was within the following temperature and time limits:

Concrete Temperature Time Limit 800 F........................ 35 minutes 700 F. . . . . . . . . . . . . . . . . . . . . . . . 4 0 min ute s 600 F........................ 55 minutes 500 F........................ 65 minutes

c. If concrete was found to be nonplastic, concrete was placed in accordance with requirements of Section 8.5.3 of ACI 301, except that no dampening was required before the application of grout.

d. All concrete was deposited in the forms after introduction of mixing water to cement and aggregates, within the following time limits:

Concrete Temperature Time Below 6 00 F................. 2-1/2 hours 610 F to 700 F............... 2 hours Above 700 F. . . . . . . . . . . . . . . . . 1- 1/ 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> B.1-13 f

1 Aceinsss l

l

. 6 B/B-FSAR AMENDMENT 42 MAY 1983 The above limitations were waived if the concrete was within the allowable limits for slump, provided it could be transported and placed without addition of water to the batch. Otherwise the concrete was rejected.

Concrete that was beyond the allowable limits for slump and air content but within the extreme limits as placed in the forms, was placed within a 1-1/2 hour time limit. Otherwise the concrete was re jec ted.

e. Hot weather concreting conformed to Section 12.3.2 of

" Recommended Practice for Hot Weather Concreting," i Section 12.3. 2 of ACI 301, except as modified I below:

i 1. Adequate provisions were made against plastic shrinkage cracking, as specified in Chapter 2 of ACI 305, " Recommended Practice for Hot Weather Concreting."

l

f. Cold Weather concreting: Cold weather concreting conformed to the following provisions:

i 1. Concrete was placed at a temperature within the allowable limite indicated in Table B.1-3 and in accordance with the minimum temperature for the l

time indicated in Table 1.4. 2 of ACI 306, I l

" Recommended Practice for Cold Weather l Concreting," and upon removal of heat, the

maximum temperature drop conformed to Table 1.4.1, Line 17 of ACI 306.

g. Where early strength was critical, as indicated by Sargent S Lundy, concrete was maintained at the minimum allowable temperatures indicated in Table B.1-3 for the period of time indicated in Table 5.1. 7 of ACI 306.

h. If construction temperature records indicated the possibility of a portion of the concrete in place being exposed to freezing temperatures prior to elapse of curing times indicated in Table 1.4.2 of ACI 306, or during placement, an investigation l with concrete test hemmer, drilled cores, or soniscope was conducted.

B.1.11 Concrete Control Tests concrete testing and sampling frequencies conformed to Tables B.1-4 and B.1-5. Allowable limits conforacd to Tables B.1-1, B.1-2, and B.1-3.

B.1-14 AOP14 mig

B/B-FSAP AMENDMENT 19 MARCH 1979 Concrete samples were obtained at truck chutes, except sa mples from pumped concrete which were obtained at point of discharge from the pump. Samples were obtained in accordance with ASTM C172, except that when a sample was secured by diverting truck chute or pipe discharge into wheelbarrow, no compositing was required; and when central mixed concrete was delivered, the sample was taken from any pcrtion of the truck discharge.

Each time sampling commenced, requirements for normal sampling as defined in Tcble B.1-4 were followed.

All normal samples were randomly selected.

When a concrete sample was taken from a truck as concrete was being discharged, discharge from the truck was immediately resumed while concrete was being tested.

If results from tests for slump, temperature, or air content were beyond the allowable limits, placement continued for the next five loads providing test results were not beyond the extreme limits and tightened sampling was instituted.

For tightened sampling, samples were taken from the next available truck whether or not its discharge had begun.

Discharge from this truck was resumed immediately af ter the sample was taken. If test results were beyond the allowable limits, discharge from this truck was discontinued and a sample was taken from the next available truck. This may have continued until five consecutive loads outside the allowable limits had been tested at which time concrete placement was discontinued until corrections were made. If two consecutive samples were tested within the allowable limits, normal sampling was resumed.

The required test specimens from each sample were molded and cured in accordance with ASTM C31.

B.1.12 Evaluation and Acceptance of Fresh Concrete Test results on f resh concrete were in accordance with the regairements in Tables B.1- 1, B.1-2, and B.1-3.

Concrete which had set was not retempered but discarded.

Concrete was rejected for remixing or wasting if any or all the following conditions existed:

a. Time limitations after introduction of water to cement were exceeded.

b. Five consecut;ve trucks or batches remained on

tightened inst.ection in Subsection B. I.11.

l l

l l

B .1- 15 AOP:b1SS7 l

B/E-FSAP AMENDMENT 43 SEPTEMBER 1983

c. Temperature, slump, or air content was beyond the B.1-1, B.1-2, and extreme limits, as listed in Tables B.1-3.

B.1.13 Evaluation and Acceptance of concrete compression Results The strength level of concrete was considered satisfactory if the following two criteria were satisfied when using the standard deviation f rom at least 30 consecutive strength tests representing similar concrete, and conditions of concrete being evaluated:

a. A probability of not more than 1 in 100 that an average of three consecutive strength tests was below specified strength.

b. A probability of not more than 1 in 100 that an individual strength test was more than 500 pri below the specified strength.

Methods in ACI 214 were used in concrete evaluation along with the above criteria.

The above criteria was considered satis fied if either:

a. The average of all sets of three consecutive strength test results at 91 days equaled or exceeded the specified compressive strength of the concrete and no individual strength test result fell below the specified compressive strength by more than 500 psi, or

b. The average compressive strength, f avo req., conformed to the following two expressions favereq. 2 fe + 1.343 o fave req. 2f c - 500 + 2.326 o where:

fc = specified compressive strength l c = standard deviation.

B.1.14 Consolidation of Concrete consolidation of concrete conformed to requirements in Section 8.3.4 of ACI 301, and the following:

a. All concrete was consolidated by sufficient vibration so that concrete was worked around reinforcement, B.1-16 AW F.SSS

B/B-FSAR AMENDMENT 42 MAY 1983 around embedded items, and into corners of forms, eliminating air or stone pockets,

b. When a layer of concrete was being consolidated, the vibrator spud penetrated at least 6 inches into the previously consolidated layer.

c. Spacing of vibrator insertions and withdrawals caused overlapping " spheres of influence", generally at about 18 inch spacing.

d. Vibrators were not used to effect horizontal movement of concrete.

e. If in the opinion of the inspector, segregation was occurring prior to adequate consolidation, adjustment of mixture or pattern of vibration was considered.

f. Internal vibrators used in the work had a minimum frequency of 8000 vibrations per minute.

B.1.15 Concrete Finishes concrete finishes for all unformed surfaces conformed to the finishes indicated on the drawings and to Sections 11.7, 11.8 and 11.9 of ACI 301 and to the following addition to ACI 301:

l

a. Section 11.7.1: Brcoming exposed some of the aggregate and scored the surface to provide mechanical bond for the separate finish.

b. Section 11.8.1: A scratch finish was used for the top of turbine and equipment foundations and top of concrete duct runs.

c. Section 11. 7. 2: Spreading of cement or a cement-sand mixture directly on top of concrete was not permitted. The finish surface was not marked off in areas or scored in any manner.

d. Section 11.8.2: A float finish was used for the to of concrete walls, floors of tunnels, crib houses, p manholes, sump pits, elevator pits, valve pits and miscellaneous pits.

e. Section 11. 8. 3: A troweled finish was used for floors, stair treads and for the top surface of curb, l piers, pads, pedestals switchyard foundations and other outdoor equipment foundations where top sur faces were exposed after completion of the work.

f. Section 11.7. 4: Strokes were square across the I surface and made so as to produce regular scoring i

B.1-17 D0145SS l

~_ _ __ _ _ _ _ .

B/B-FSAR AMENDMENT 42 MAY 1983 without tearing the surface or exposing aggregate.

Scoring ran transverse to the direction of traffic.

g. Section 11.8.4: This finish was used for driveways.

This finish was used for other surfaces only if specifically indicated.

For floor surfaces covered with chemical-resistant floor the finish surface was dropped 1/4 inch so that the chemical-resistant finish was flush with adjacent floor areas.

B.1.16 Curing and Protection Curing and protection conformed to the requirements of Chapter 12 in ACI 301 and the following: l

a. Subsections 12. 2.1.1 through 12. 2.1.6 and 12. 2. 2 of ACI 301 did not apply.

b. Where forms were stripped before completion of specified curing period, curing compound was applied immediately after completion of specified surface treatment.

c. Prior to initial use of specified compounds, the manufacturer's technial representative visited the jobsite and personally gave instructions on the correct use of materials,

d. To control membrane thickness, compounds were applied at the rate of approximately 300 ftr/ gal, unless otherwise instructed by manufacturer.

e. Curing was continued for not less than the minintan periods specified in Section 12. 2.3 of ACI 301 before applying any other surfacing or before opening l to traffic.

l B.1.17 Preplaced Aqqregate Concrete Preplaced aggregate construction conformed to provisions of Chapter 7, "Preplaced Aggregate Concrete" of ACI Standard 304,

" Recommended Practice for Measuring, Mixing, Transporting, and Placing Concrete," and to the following:

a. Time of efflux for the premixed grout was in the range of 20 to 24 seconds when measured immediately after mixing in accordance with CRD-C79, " Flow of Grout Mixtures (Flow-Cone Method) ," and

b. The preplaced aggregate concrete test specimens made l in accordance with CRD 84 attained a strength of 5500 t

psi in 91 days B.$-18 goeMSSO

• I B/B-FSAR AMENDMENT 42 l MAY 1983 B.1.18 Evaluation and Acceptance of concrete The evaluation and acceptance of concrete was accomplished in conformity to Chapters 17 and 18 of ACI 301 l

a l

B.1-19 A')D3.4591 4

,- _ , - ,,- _ , , - - . - . - . ~ . , . _ - , . . . _ . - . . . . _ - - - -

1 I

TABLE B.1-1 AIR CONTENT l

I ALLOWABLE LIMITS 1 COARSE AGGREGATE TOTAL AIR CONTENT (VOL.) % ,

NOMINAL MAXIMUM

! SIZE IN COARSE FREEZING AND TIIAWING FREEZING AND THAWING

. ASTM C 33 AGGREGATE (in.) RESISTANCE REQUIRED RESISTANCE NOT REQUIRED Allowable Extreme Allowable Extreme l Limits Limits Limits Limits i

8 3/8 7 to 9 6 to 10 3 to 9 3 to 10 m 67 3/4 5 to 7 4 to.18 2 to 7 2 to 8 E

o 57 1 4 to 6 3 to 7 1.5 to 6 1.5 to 7 y N

I

' nG G *9 i 3 WG

! h 0"~

c: *

'J N

I i

i

~

/B FSAR M1ENDMENT 19 MARCH 1979 TABLE B.1-2 LIMITS FOR SLUMP CONCRETE TEMPERATURE ALLOWABLE LIMITS EXTREME LIMITS AS PLACED ( F) (in.) (in.)

Minimum Maximum Minimum Maximum Below 55 2 5 1 6.0 Between 55 and 64 2 4.5 1 5.5 Between 65 and 74 2 4 1 5.0 Between 75 and 85 1.5 3.5 1 4.0 l

i l

l i

i A003.4593 l B.1-21

TABLE D.1-3 PLACING TEMPERATURE ALIDWED LIMITS FOR CONCRETE EXTREME VALUES FOR CONCRETE TEMPERATURE AS PLACED (* F) TEMPERATURE AS PLACED (* F )

Exposed concrete face (s) normal Moderately Moderately to the thickness Thin Massive Thin Massive of the pour Section Section Massive Section Section Massive One face exposed $12 12 to 48 >48 312 12 to 4C >48 l Two opposite ,

faces exposed $18 18 to 72 >72 <18 18 to 72 >72 l s

L Between Max. Max. Max. Max. Max. Max.

4, 90 and 81 80 75 70 u

to a: s 85 80 75 lR in 2$ Between Max. Max. Max. Max. Max. Max. 43

%@ 80 and 46 90 80 70 90 85 75 OO l$

u Max.

lc Between Max. Max. Max. Max. Max.

g 3 @ ,_

f 90 80 75 90 85 80 8-- 45 and 26 Min. Min. Min. Min. Min. Min. l

, I@

to o 55 50 45 50 45 40 kO Between Max. Max. Max. Max. Max. Max.

N$ "

90 80 75 90 85 80 @$

25 and 0 Min. Min. Min. Min. Min. Min.

60 55 50 55 50 45 ! =yay

> wk e *4

= w a

Cil t

i 44

B/B-FSAR AMENDMENT 42 MAY 1983 TABLE B.1-3 (Cont ' d )

Notes:

1. No concrete was poured when surrounding air in contact with the concrete was below O' F.

2. In all cases subsequent freening of concrete was prevented by providing the protection recommended l

in Table 1.4.2 of ACI 306.

3. Since metal deck and noninsulated formwork do not prevent heat dissipation significantly, concrete surfaces in contact with then were considered as having exposed faces.

4. When concrete was placed at a temperature exceeding 70* F, cement was added and mix adjusted if water-cement ratio exceeded that of the mix design. In computing water-cement ratio, total water available as mixing water in concrete from whatever source was considered. Adjusted mix proportions, including total water available, were shown in the inspector's report and were reported with the strength test results.

B.1-23

AOP3 ASS 5

T TA.L3 S. :-e rY100C.FT. (Y.tP.P.9 f nes itsTI.G e.rui .cv raram, i canc.o., s i

,3, g co nenime.v. man

.ue m .ner .m...: r..t.

c. .

.i., .:

m.. -- s, m ..

M h.f or c n un , e,.ti.,

r.

n~.. .: .m, ,,,m. Ibu.,h,er

c. ,m o. e r.a,ts.,

- m ,. of s

jU5

.g c.

c, .pe u. a asuu c si sie ..en reo.

si.

rio.i.s.e.n s ,a

..a.

v....a .e

v. as . on.

..= ,ne ..en es =

si era.i.siwm s s

.is a pt. .r. . e .a t

i n. .en

. . . , z celere si e , o. n..icn . .,s.,3.

i e.a t....a

.. 2 I

- .nc, se.e.c.ase.v cn 4 . es a.r a resa.ee.ase case .a 7. as as *w

,a comp ... c :o ru.a ..e ar i... es c ..en 4. . v.'.

ie i e. , .' i ,r .ie .e a.

.. =.e. -.a...

a.r *. *i 4.,.

• ser ,em pr.e in . sw case rua. .6.. a.o ca.e v., a.

.[ .m see e.nie rua.

g;I

• si r.. .a . ::: ..c fr 4. .a t. is

m. ,::, ..cn reo. si.

e. i.si.e,.,.ir.a v....a..e on.

...< ,in ase..cn rio. i.s i .. .,,. a.ac.a .m.w, ano cose si

r. .n.i.. ,.,.... .. s.

g

.: cyt i.au ca

... c si ..u-a . see c e se is ..pi.

7 n w

+i a.r. v - a-

< *ie ser. i.. . is. pi.

u v. n a r. e. s*=*i. is...pa.

,2.-'.

E.23 s B. .

s.~...i..

a. .s. ie. a. .s.o. . tc. 4.i .' i a. .s.s. .tc . : a 4.f .

r. .

2.

- g3 -."

g g; co. en w

i.a i . .,. t . .a s.e o.r ,i..w ,osa. pa r.. .a, et .

v i .a .i .,.

e.r.o.

s

,y . e4 pa. : . pi.

. .a si a.y c i

i.a i ... . .ss.

s.o,. i..uca y .o.pi. .r..

i .

4.y.

a f, E

3 pg 3 -= si.

in i.i.s.v. is .. 1. i.s. v. .ie. n . s.s. t. ec. i

=

a I

m. in nem rios

, wry see cale si

r. i.s: e .,. i. .a v..e.a..

v. as .

on.

.. ,o n ..en e, si <

r .ei

.w.., ir..a r.. .4

.e 7 as 2n. ::: ..c. ri o.

...,, no ease sreo.i . n..i ... .,..ini,a

. v...

. 1

.a rua. ui s..

,emi.sa se pa.

ag 2.I 7cr aan.

c si m.n e =.

s. ,1

..p a s . . .. . t a . ie.

et a.r. a vna iso casc n.+ s.na. a.v.

is. pa ee.: en .na

.w

• ru a. s .p i. i nar i.

s.....u. .te.

ae w e 4.r.

,3 'g g;i.

2  ::: v r....a t...

• • f om r. . .i i,.a r..t.a 9: ay sa.i i,r.a 9: d.p er n i se .,u., .t b Do pr 54 C 29 s.r.,se  : .. pa. r .pse s.e o. t . . a4 ..i ..gp i. s.epie. .a.ai ts eep ie 4.y.i

$$" Bts gth i.a. .s.7.e. etc. : i. a. . s. 7.e, .te. i .a. . s. 7.8

.tc.

o N

53 mansermrwm .... ..- . i . .w ~.. -.. : . . .. . . u . u a a-. .

IN

/

's*

G i -e b

.Cnm M

TABLE B.1-5 CONCRETE TESTING FREQUENCY CATEGORY I - CATEGORY II TYPE TEST ASTM CONTAINMENT

• OTHERS Slump C 143 3 First batch placed each For each 50 yd of concrete, or for3each Air Content C 173 day and for each 50 yd 3 day's placement, if less than 50 yd .

C 231 placed.

ai$ Temperature C 138 Daily during production Not Required  !

fU o ,1 Unit Weight /

yield ca z2

" C 94 Initially and every 6 months Not Required Mixer u uniformity to

.t" u

N d, Slump C 143 When tests on Normal Samples showed measurement of a concrete property, I m On temperature, slump, or air content out of allowable limits but within 3 un g gz the extreme values, an Additional Sample was taken from chute of the >

a: F. a Air Content C 173

• b C 231 If measurement of this additional sample showed 1

~ =c next available truck.

this property to be within allowable limits, and deviations were not

" " directly attributable to the transport from the truck chute to the forms, Temperature this truck load was placed. If not within allowable limits a second Additional Sample was then taken from the next available truck and tested.

This procedure was continued until tests on two successive Additional Samples had indicated that the concrete properties were within allowable limits. Normal sampling was then resumed.

G Reactor cavity, tendon tunnel, and containmert basemat, shell, and dome.

Ek Eg

" External Concrete l zo M. x t's c,

U" l 9 i

i a

B/B-FSAR AMENDMENT 19 MARCH 1979 l TABLE B.1-6 GRADATION OF HEAVYWEIGHT AGGREGATE Fine Aggregate:* Spec.

Required Sieve Size 3/8 in. 100

1. 4 75 - 95 98 55 - 85 916 30 - 60 l

1. 30 15 - 45

1. 50 10 - 30

1. 100 5 - 15 Coarse Aggregate:* Spec.

l Required Sieve Size

! 1 in. 100 3/4 in. 90 - 100 i 1/2 in. --

3/8 in. 20 - 55

1. 4 2 - 15

1. 8 0-8 e

i j

• Fine Aggregate: 10% of the material passing the 3/8-inch sieve was allowed to pass the No. 200 sieve if the material

! passing the No. 200 sieve was shown to be essentially free of clay or shale.

AO?S.1599 B.1-26

E/B-FSAR AMENDMENT 42 MAY 1983 1

B. 2 REINFORCING STEEL l B.2.1 Requirements for Category I Materials l Reinforcing bars for all Category I structures were Grade 60 deformed bars tested in accordance with criteria in NRC Regulatory Guide 1.15 for " Testing of Reinforced Bars for Category I Concrete Structures. " They met the requirements of ASTM A615, " Specifications for Deformed and Plain Billet-Steel l l Bars for Concrete Reinforcement," with the following modifications. Paragraphs 4.2, 7.3, 8.3 and all of Sections 9 and 10 as indicated below were used in lieu of th? same parts as specified in A615

a. 4. 2 The chemical composition thus determined was transmitted to Purchaser or his representative.

b. 7.3 The percentage of elongation for bars Nos. 3 through 11 was as prescribed in Table 2.

1

! c. For bars Nos. 14 and 18, the minimum elongation in 8 inches (full-section specimens) was 125.

l

d. 8.3 Bars of size Nos. 14 and 18 were bend tested as required below in Section 9.3.

i e. 9. Test Specimens

-l j 9.1 All tension test specimens were full-section of

the bar as rolled and randomly sampled.

9.1.1 The test procedures were in accordance with ASTM A-370, " Methods and Definitions for l Mechanical Testing of Steel Products."

i 9.1.2 Delete.

9.2 The unit stress determinations on full size specimens were based on the nominal bar cross-sectional area as in Table 1.

! 9.3 The bend test specimen was full-section of the i

bar as rolled. The pin diameter for the 90* Bend r Test was equal to 10d. for Bars Nos.14 and 18. l l f. 10. Number of Tests.

10.1 At least one specimen from each bar size was tested for each 50 tons or fraction thereof, of the reinforcing bars that were produced from each heat.

B.2-1 AM.%S9

B/B-FSAR AMENDMENT 42 MAY 1983 10.2 Testing included both tension tests and bend tests.

10.3 If any test specimen developed flaws, it was discarded and another full-section specimen of the same size bar from the same heat substituted.

10.4 If any of the tensile properties of one out of the total number of test specimens corresponding to a heat was less than that specified in Section 7 as modified herein but was greater than the limits shown below, retest was allowed:

Grade 60 Tensile strength, psi 83,000

, yield stress, psi 55,000 i

Elongation in 8 inches, percent Bar no.

3, 4, 5, 6 6 7, 8, 9, 10, 11 5 14, 18 9 10.a.1 The retest consisted of at least two additional full-section tensile tests on samples of l the same bar size and heat fraction.

10.4.2 Each one of the additional test specimens and the average of all of the test specimens corresponding to the same bar size for this heat, (including the original one) met the requirement of

Section 7 of ASTM A615, as modified herein.

10.5 If the original test failed to meet the limits indicated in Paragraph 10.4, or if any tensile or bending property of specimens retested in accordance with Paragraphs 10.4.1 and 10.4.2 did not meet the requirements of Section 7 of ASTM A615, as l modified herein, that material was rejected.

10.6 If any tensile property of the tension test specimen was less than that specified in Section 7 of ASTM A615 ,as modified in Paragraph 10.4, and any an part of the fracture was outside the middle third of the gauge length, as indicated by scribe scratches marked on the specimen before testing, a retest was permitted.

i All reinforcing was tagged or marked in a manner to ensure j traceability to the certified Material Test Report (CMTR) during l production, fabrication, transportation and storage.

B.2-2 A0014600

B/B-FSAR AMENDMENT 42 MAY 1983 Traceability for all reinforcing bars was by the original heat number.

Traceability of all reinforcing was completed up to the placing of the reinforcing, which was considered as the last hold point for the bars.

B.2.2 Reinforcing Bar Fabrication Fabrication for all reinforcing bars conformed to the requirements in Chapter 7 of CRSI " Manual of Standard Practice" and to the following:

a. Bar ends for bars which were spliced using Cadweld procedures were checked for clearance after shearing, using a test sleeve with a star.dard Cadweld sleeve.

B.2.3 Cadweld splicing Splices in rein forcing bar sizes No. 11 and smaller here lapped  ;

in accordance with ACI 318, " Building Code Requirements for l Reinforced Concrete," or Cadweld spliced. Bar sizes No. 14 and No. 18 were Cadweld spliced. The splice was designed to develop the specified minimum ultimate strength of the reinforcing bar.

B.2.3.1 Oualification of Ocerators Prior to production splicing, each Cadweld operator prepared two qualification splices for each position used in his work. These were tested and met the joint acceptance standards for i

workmanship, visual quality, and minimum tensile strength.

B.2.3.2 Procedure Specifications All joints were made in accordance with the manufacturer's

instructions, "Cadwell Rebar splicing," plus the following additional requirements

a. A manufacturer's representative, experienced in Cadweld splicing of reinforcing bars, was required to be present at the jobsite at the outset of the work

, to demonstrate the equipment and techniques used for i making quality splices. He was present for the first 1

25 production splices to observe and verify that the l equipment was being used correctly and that quality splices were being obtained. The Cadweld manufacturer furnished the Certified Material Test Report for each lot of splice sleeve material delivered. This report included the physical and chemical properties of the sleeve material. The splice sleeves, exothermic powder, and graphite molds were stored in a clean dry area with adequate B.2-3 A0014601

B/ E-FS AE AMENDHENT 19 MARCH 1979 protection from the elements to prevent absorption of moisture.

b. Each splice sleeve was visually examined immediately prior to use to ensure the absence of rust and other foreign material on the inside diameter surface, and to ensure the presence of grooves in the ends of the splice sleeve.

c. The graphite molds were preheated with an oxyacetylene or propane torch to drive-of f moisture at the beginning of each shift when the molds were cold or when a new mold was used.

d. Bar ends to be spliced were power-brushed to remove i all loose mill scale, loose rust, concrete, and other i foreign material. Prior to power-brushing, all

! water, grease, and paint were removed by heating the bar ends with an oxyacetylene or propane torch.

e. A permanent line was marked 12 inches back from the end of each bar for a reference point to confirm that the bar ends were properly centered in the splice sleeve. In those cases where the 12 inch gauge length was not practical, different gauge lengths were used, provided they were properly documented.

f. Immediately before the splice sleeve was placed into final position, the previously cleaned bar ends were preheated with an oxyacetylene or propane torch to j ensure complete absence of moisture.

l 1 g. Special attention was given to maintaining the alignment of sleeve and pouring basin to ensure a proper fill.

l h. The splice sleeve was externally preheated with an i

oxyacetylene or propane torch after all materials and equipment were in position. Prolonged and l unnecessary overheating was avoided.

i

1. Each splice was examined by the operator prior to forming to ensure compliance with all requirements.

All completed splices and sister test specimens were stamped with the operator identification mark.

B.2.3.3 Visual Examination All completed splices (including the sister test specimens) were inspected to ensure compliance with the visual examination acceptance standards. Splices that failed any requirement were rejected and replaced and not used as tensile test samples.

e.2-4 A0014602

B/B-FSAR AMENDMENT 19 MARCH 1979 i l

All visual examinations on completed splices were performed only after the splices had cooled to ambient tempe ra ture. The visual examination acceptances standards were:

a. Filler metal was visible at the end(s) of the splice sleeve and at the tap hole in the center of the sleeve. Except for voids, the filler metal recession was not more than 1/2 inch from the end of the sleeve.

b. Splices did not contain slag or porous metal in the tap hole or at the end (s) of the sleeves. When in doubt as to whether filler metal or slag was in the tap hole, the riser was broken with a punch or file, filler metal shines while slag remains dull. If slag was f ound, the Inspector removed slag at the tap hole and searched for filler metal. This requirement was not cause for rejection unless the slag penetrated beyond the wall thickness of the sleeve.

c. A single shrinkage bubble present in the tap hole was distinguished f rom general porosity and it was not cause for rejection.

d. The total void area at each end of the sleeves did not exceed the following limits (for splicing bars up to Grade 60):

1. for No. 18 bars - 2.65 inz

2. for No. 14 nars - 2.00 in2

3. for No. 11 cars - 1.5 ina 4 for No. 10 bars and splice Catalog Number RBT-10101 (H) 1.58 in2

5. for No. 5 bars - 0.53 in ,z

e. The distance between the gauge lines for a type "T" splice was 24-1/4 inches i 1/2 inch for the 12 inch gauge lengths, or (X + Y + 1/4) i 1/2 inch when the X and Y gauge lengths are used. The center of the gauge line connecting the gauge marks fell within the diameter of the tap hole,

f. The distance between the gauge line and the structural steel for Type "B" splice was 12-1/4 inches t 1/4 inches or any other documented distance.

l l

i 2.2-5 g>M603

i B/B-FSAR AMENDMENT 42 l MAY 1983 B.2.3.4 Samplina and Tensile Testing Splice samples were production splices and sister splices.

Production splice samples were not cut from the structure when Type "B" splices were used, or when Type "T" splices were used for curved reinforcing bars. Representative straight sister splice samples were used in such cases, using the same frequency as Type "T" splices on straight bars, except that all splice samples are sister splices. Separate sampling and testing cycles were established for cadweld splices in horizontal, vertical, and diagonal bars, for each bar grade and size, and for each splicing operator as follows:

a. one production splice out of the first ten splices,

b. one production and three sister splices for the next

ninety production splices, and

c. one splice, either production or sister splices for the next subsequent units of 33 splices. At least 1/4 of the total number of splices tested were production splices.

The tensile testing acceptance standards were:

a. The tensile strength of each sample tested was equal or exceeded 125% of the minimum yield strength specified in the ASTM A615 for the grade of reinforcing bar using loading rates as stated in ASTM A370 for the grade of reinforcing bar.

b. The average tensile strength of each group of 15 consecutive samples was equal to or exceeded the ultimate tensile strength specified in ASTM A615 for the grade of reinforcing bar. l Procedure for Substandard Tensile Test Results:

a. If any production splice used for testing f ailed to .

meet the strength requirements in (a) above and failure did not occur in the bar, the adjacent production splices on each side of the failed splice were tested. If any sister splice used for testing failed to meet the strength requirements in (b) above and failure did not occur in the bar, two additional sister splices were tested. If either of these retests failed to meet the strength requirements, splicing was halted. Splicing was not resumed until the cause of failures were corrected and resolved to the satisfaction of Sargent & Lundy.

b. If the running average tensile strength indicated in (b) above failed to meet the tensile requirements B.2-6 l A0014604

B/ E-FS AR AMENDME.NT 19 MAPCH 1979 stated therein, splicing was halted. Sargent & Lundy investigated the cause, determined what corrective action (if any) was necessary, and notified the Contractor to perform the corrective action (if any).

c. When mechanical splicing was resumed, the sampling procedure was started anew.

l e

B.2-7 '

A003.4605 l 5

B/B-FSAR AMENDMENT 42 MAY 1983 B.3 POST-TENSIONING TENDONS B.3.1 Ge ne ral t

A Birkenheimer, Brandestini, Poss, and Vo@ (BBRV) post-tensioning system was used. Tendons consisted of 170 1/4-inch diameter parallel lay wires. Positive anchorage at ends was

. provided by buttonheading. The materials, erection and fabrication procedures, and testing requirements conformed to the technical provisions of Sections cc-2400, cc-4400, and cc-5400 of the 1973 ASME BSPV Code,Section III, Division 2, Proposed Standard Code for concrete Reactor Vessels and Containments, issued for interim trial use and comments with the exception of CC-4464. l

] B.3.2 Materials

, B.3.2.1 Tendon Material The 1/4-inch diameter wire conformed to cold-drawn ASTM A421, l Type BA, stress-relieved, having a guaranteed minimum ultimate tensile strength, (f u, f 240,000 psi and a minimum yield strength not less th8n)0.80 f

extension under load method. Eg, as measured by the 1.0%

I B.3.2.2 Button heads i

The positive anchorage of tendons to anchor heads was provided by buttonheading of the wires. All buttonheads were cold-formed after threading wires through wire holes of anchor heads.

Buttonheads were formed symmetrically about the axis of wires and

, were free from harmful seams, fractures, and flaws.

B.3.2.3 Tendon Sheathinq l

I Tendon sheathing through the foundation consisted of black seam-

)

' less steel pipe, ASTM A53, Grade B and the wall sheathing was a black interlocked steel strip conduit, 22 gauge minimum wall thickness, fabricated to be watertight. The inside diameter of the sheathing was approximately 4.75 inches. All splices were

} sealed to prevent intrusion of cement paste. The tendon sheath splice was made using a snug fitting coupling approximately 1 foot long. The joints between the sheath and the coupling were i

taped. The minimum radius of curvature used was 30 feet, except j

in certain be acceptable. cases where a smaller radius of curvature was shote to 4

B.3.2.4 Permanent Corrosion Protection i'

A corrosion preventing grease, Viscono Rust 2090 P-4, Nuclear Grade was used as a tendon casing filler.

B. 3-1 g 03.160G

. . . ~ . - . ..,,---n.--- - , , , .

B/B-FSAR AMENDMENT 42 MAY 1983 B.3.2.5 Anchor Heads The anchor heads conformed to ASTM A-322," Specifications for l l Hot-Rolled Alloy Steel Bars," AISI 4140/4142 hot rolled, vacuum )

i degassed, and heat treated to Rc 42 i 2 per MIL-H-6875D with a guaranteed maximum annealed hardness of 217 Brinell.

B.3.2.6 Bearino Plates and Ships i i

Bearing plate and shim raterials conformed to hot rolled ASTM A-36 plate to silicone-kilned fine-grain practice.

B.3.3 Quality control

} B.3.3.1 Testin_g The erection and fabrication procedures conformed to Section

cc-4400 with the exception that the welding procedures and welder qualifications were in accordance with AWS D1.1.

I B.3.3.1.1 Tendon Tests Tensile tests were performed cn 100-inch long samples taken at a i rate of 15 of all tendons including all anchorage hardware. One l test was performed on the vertical group, one test on the dome

! group and two tests were performed on the horizontal group. The ,

tendons were required to carry a load corresponding to 1005 of the guaranteed ultimate tensile strength of the tendon without I

failure. Failure of any anchorage component was unacceptable.

B.3.3.1.2 Tests on Wires and Buttonheads Wires were tested in accordance with ASTM A421, "Specifica-l tions for Uncoated Stress-Relieved Wire for Prestrased j concrete." A bend test and buttonhead test was made on each coil 4

o f wire. The bend test specimen was cold bent back and forth in

one place 90* in each direction over pins with a 5/8-inch radius.

Each return to vertical was one complete bend. The wire must have sustained a minimum of six bends before complete fracture.

The buttonhead test was a static test performed to check the l buttonhead machine and to confirm the integrity of the buttonhead. The buttonhead was acceptable if failure occurred

! within the shaf t of the wire. The buttonhead machine was i routinely checked at the beginning of each shift. Ten percent ot the wires in each tendon were checked for buttonhamo size and conformance with a nGo, No-Goa gauge. All buttonheade were visually examined to ensure that splits, cracks, and/or slips did not exceed acceptance criteria. Also, one rupture test of wire was performe(. in the field similar to that used in the tendon. A i

12-inch long sample of wire was tested using a portable tensile test apparatus prior to initiating the buttonheading operation on B . 3-2 10014607

B/B-FSAR AMENDMENT 42 MAY 1983 the respective tendon. The sample was prepared using the same i equipment and operators that performed the buttonheading l operation.

B.3.3.1.3 Tests on Corrosion Preventative Grease -

The manufacturer of corrosion preventative tendon coating materials performed chemical analyses to measure the presence of water soluble chlorides, nitrates, and sulphides and provided certification of compliance with the acceptance criteria given in the ASME Code,Section III, Division 2. In addition, each shipment of permanent corrosion preventative grease was retested in the field to verify that the material had remained contaminant free.

B.3.3.1.4 Anchorage Hardware Tests and Inspection i

Anchor heads were tested to 120% of the minimum ultimate tensile l strength of the prestressing steel employing a test machine that was chosen to simulate the actual loading condition as close as possible. All welds were given 100% visual examination for completeness, workmanship, and slag removal.

B.3.3.2 Fabrication Tolerances I

The differential length of any two wires in the same tendon did not exceed 1/16 inch for wires up to 100 feet long and 1/8 inch for wires over and up to 200 feet long, and an additional 1/8 inch for each 100 feet increment in length over 200 feet.

I Trumpet perpendicularity was measured to ensure that the angle between the trumpet and the bearing surface of the bearing plate was within a tolerance of + 0.3 degrees. Eccentricity of a buttonhead from the axis oY the wire was not permitted to exceed 0.010 inch. Wire holes in anchor heads must have been within 0.010 inch of the specified location on the buttonhead bearing surface. Drift was within 0.035 inch from the centerline. Wire hole diameter must have fallen within the range of 0.257-inch to 0.264 inch S.3.3.3 Field Installation Tolerances Tendon bearing plates were installed with a tolerance of + 0.25 inch from the specified locations. Critical dimensions were

! established for the placing of tendon sheathing and the tolerance on these critical dimensions was + 0.5 inch. All gauges, instruments and jacks were calibrated against known standards that l were traceable to the National Bureau of Standards. Elongation ,

measurements commenced at 20% GUTS. The tolerance on lockoff pressure during the stressing operation was established by the criteria that stress in the tendon wires at the anchor point after anchoring must have been at least equal to, but could i not have exceeded, the specified value by more than 5%. The i number of broken or defective wires or buttonheads was limited to l a maximum of three per tendon. The total number of broken or l defective wires in any one group of tendons (hoop, vertical, or B.3-3 AW3.4608

3/B-FSAR AMENOMENT 19 MARCH 1979 dome) was not allowed to exceed It of the total wires in the tendon group.

B.3.3.4 corrosion Protection Tendons were protected from corrosive elements during fabrication, shipping, storage, and installation by application of a thin film of Visconorust 1601 Amber, as made by Viscosity oil company, immediately after fabrication. Further, tendons were shipped and stored in polyethylene bags. Tendons were not permitted to be exposed to inclement weather, condensation, or injurious agents such as soluticns containing chlorides. Damaged or corroded tendons were rejected on inspection. Exterior exposed surfaces of bearing plates and grease retaining caps were protected from corrosion by application of a prime and a finish coat of paint. The prime coat was DeGraco 501 and the finish coat was DeGraco 30, as made by the Detroit Graphite Company.

The prime coat of paint was required to have a minimum dry film thickness of two mils.

I 1

4 l

B.3-4 A0('14609

B/B-FSAR AMENDMENT 42 MAY 1983 B.4 STRUCTURAL STEEL B.4.1 Structural Steel Materials Structural support steel was ASTM A36, ASTM A572, Grade 50 and ASTM A588 high strength, low alloy corrosion-resistant steel.

Structural steel tubing was ASTM A500, Grade B and ASTM A501.

B.4.2 Structural Steel Connections and Connection Material B.4.2.1 Bolted Connections Structural steel bolted connections used ASTM A325, " Specifications l for Structural Steel," Type 1 and ASTM A490, " Quenched and Tempered Alloy Steel Bolts for Structural Steel Joints," friction-type high strength bolts. These high strength bolted connections l conformed to " Specification for Structural Joints using ASTM A325 or A490 Bolts" issued by the Research Council on Riveted and Bolted Joints of the Engineering Foundation and endorsed by the AISC, and to Framed Beam Connections, Table I or II of the AISC Manual. ASTM A307 and A325 bolts were used for non-friction type applications in specified connections in the containment building. For non-friction type connections these bolts were tightened to a specified torque range. ASTM A36, " Specifications for Structural Steel," nuts were used with ASTM A36 threaded rods and all ASTM A307, " Specifications for Carbon Steel Externally and Internally Threaded Standard Fasteners," bolts. l B.4.2.2 Welded Connections Standard welded beam connections conform to Table III or IV of AISC Manual. Shop and field welding procedures were in accordance with AWS Specifications listed in Table 3.8-2.

Selection of electrodes and recommended minimum preheat and interpass temperature were in accordance with AWS requirements.

All welders and welding operators were certified by an approved testing laboratory and were qualified under AWS procedure as stated in AWS Specifications.

B.4.3 Quality Control B.4.3.1 General Quality assurance requirements applied to the fabrication and testing of structures and components. Certified material test reports were furnished stating the actual results of all chemical analyses and mechanical tests required by ASTM specifications.

Identifying heat numbers were furnished on all structural steel to trace the steel to the specific heat in which the steel was made.

B.4-1 A')tt14G10

. o B/B-FSAR AMENDMENT 42 MAY 1983 B.4.3.2 Testing and Inspection of Weldments One hundred percent of all complete penetration groove welds had complete radiographic examination, except that welds impractical to radiograph were examined by ultrasonic, magnetic particle, or liquid penetrant methods.

The above nondestructive test methods were in compliance with the following ASTM specifications:

a. E94, " Recommended Practice for Radiographic l Testing,"

b. E142, " Controlling Quality of Radiographic l Testing,"

l c. E164, " Recommended Practice for Ultrasonic l Contract Examination of Weldments,"

d. E109, " Dry Powder Magnetic Particle Inspection,"

l

c. E138, " Wet Magnetic Particle Inspection," and I f. E165, " Recommended Practice for Liquid Penetrant l l Inspection Method." l B.4.3.3 Fabrication The fabrication of structural steel conformed to AISC specifications.

i l B.4-2 'N 'E U

I . .

i B/B-FSAR AMENDMENT 42 MAY 1983 l B. 5 CONTAINMENT LINER WITHIN THE CONTAINMENT BACKED BY CONCRETE

B. 5.1 General, The materials, erection and fabrication procedures, and testing I

, requirements conformed to the technical provisions of Sections CC-2500, CC-4500, and CC-5500 of the 1973 ASME BSPV Code,Section III, Division 2.

B.S.2 Materials The containment liner materials performing only a leaktight i function (excluding leak test channels), within the containment backed by concrete met the requirements of the ASME B&PV l

} Code,Section III, Division 2, Paragraph CC-2500, and complied with the following specifications:

APPLICATION SPECIFICATION Liner Plate SA 516 GRADE 60 l l Containment Liner Anchors A36 l 3

B.5.3 Quality Control B.S.3.1 Testino of Welds B.S.3.1.1 General

! All nondestructive examination procedures were in accordance with l Section V of the ASME B&PV Code.

l B.S.3.1.2 Liner Plate Seam Welds i

B.S.3.1.2.1 Radioaraphic Examinations l

The first 10 feet of weld for each welder and welding position was 1005 radiographed. Thereafter one spot radiography of not less than 12 inches in length was taken for each welder and I welding position in each additional 50 foot increment of weld.

In any case a minimum of 25 of liner seam weld was examined by

radiography. All radiographic examinations were performed as l soon as possible after the weld was placed. The spots selected l for radiography were randomly selected. Any two spots chosen for i

radiographic examination were at least 10 feet apart. If a weld failed to meet the acceptance standards specified in NE-5532,Section III of the ASME BSPV Code, two additional spots were

examined at locations not less than 1 foot from the spot of initial examination. If either of these two additional spots failed to meet the acceptance standards then the entire weld test unit was considered unacceptable. Either the entire unacceptable weld was removed and the joint rewelded, or the entire weld unit B.5-1 1003.4612

- . . = .- .- - _ . .

B/B-FSAR AMENDMENT 42 MAY 1983 was completely radiographed and the defective welding repaired.

The repaired areas were spot radiographed.

B.5.3.1.2.2 Ultrasonic Examinations 1

Ultrasonic examinations were performed on 100% of the jet  ;

deflector support embedments. If a weld failed to meet the i acceptance standards specified in NE-5330 of Section III of the ASME BSPV Code, the weld was repaired and reexamined.

B.S.3.1.2.3 Magnetic Particle Examination Magnetic particle examination was performed on 100% of liner seam welds for ferritic material. If a weld failed to meet the acceptance standards specified in CC-5533 of Section III of the ASME B&PV Code, the weld was repaired and reexamined l according to the above Code using magnetic particle examination.

B.S.3.1.2.4 Liquid Penetrant Examination Liquid penetrant examination was performed on 100% of liner seam welds for austenitic materials. If a weld failed to meet the acceptance standards specified in CC-5534 of Section III of the ASME B&PV Code, the weld was repaired and reexamined l according to the ASME Code using the liquid penetrant method of examination.

B.5.3.1.2.5 Vacuum Box Soap Bubble Test The vacuum box soap bubble test was performed on 100% of liner seam welds for leaktightness. If leakage was detected the test i was repeated after the weld was repaired.

B.S.3.1.3 Leak Test Channels Wherever leak-chase-system channels were installed over the liner welds, the channel-and-liner plate welds were tested for leaktightness by pressurizing the channels to the containment design pressure and doing a pneumatic test of 100% of the welds.

A 2 psi change in pressure over a 2-hour holding period was allowed because of a possible variation in temperature during the holding period.

B.5.3.2 Fabrication and Installation B.S.3.2.1 General The fabrication and installation of the containment steel boundaries backed by concrete were in accordance with the ASME BSPV Code,Section III, Division 2, Paragraph CC-4500.

B.5-2 A003AG13

~ _ _ - , . _

_ . _ _ . _ - _. - _ . ._ ~

B/B-FSAR AMENDMENT 42 MAY 1983 B.S.3.2.2 Weldino cualification The qualifications of welders and welding procedures were in accordance with Section III, Division 2, Paragraph cc-4500 of the i

ASME B&PV Code. l Installation Tolerances '

All pressure retaining components conformed to the applicable requirements of NE-4220 of ASME Section III.

cylinder Tolerances:

a. For each 10 foot elevation of the liner the difference between the maximum diameter and minimum diameter did not exceed 8 inches. This requirement was satisfied by measuring diameters spaced approximately 300 *

b. The radius of the liner was within i 6 inches of the theoretical radius.

c. The deviation of the liner from true vertical did not i exceed 1 inch in any 10 feet nor 3 inches in the full height of the liner.

i

d. The local contour of the shell was controlled by limiting the following deviations:

1. A 1-inch gap between the shell and a 15-foot-long template curved to the required radius when placed against the surface of a shell within a single plate section and not closer than 12 inches to a welded seam.

2. A1 1/2-inch gap when the template above was placed across one or more welded seams.

3. A 3/8-inch gap when a 15-inch-long template curved to the required radius was placed against the surface of the shell within a single plate section and not closer than 12 inches to a welded seas.

4. A 3/4 inch deviation from a 10-foot straight edge placed in the vertical direction between circumferential seams.

Dome Tolerances:

a. For each point the height of the dome above the spring line was no greater than 12 inches above theoretical height but in no case was it less than the theoretical height above the spring line.

I 8 5-3 A0014614

• O B/B-FSAR

b. Radius measurements were taken at the top of each roof course at 300 intervals, to determine the horizontal distance from the vertical centerline of the containment to the dome roof liner plate.

c. The local contour of the dome was controlled by limiting the following deviations:

1. A 1-inch gap hetween the shell and 4 15- foot-I rng template curved to the required radius when placed horizontally against the surf ace of the shell within a single plate section and not closer than 12 inches to a velded seam.

2. A1 1/2-inch gap when the template above was placed horizontally across one or more welded sears.

3. A 3/8-inch gap when a 15-inch-long template curved to the required radius was placed horizontally against the surface of the shell within a single plate section and not closer than 12 inches to a welded seam.

4. A 3/8-inch gap when a 15-inch-long elliptical template was placed along the meridional of the surface of the shell within a single plate section and not closer than 12 inches to a welded seam.

5. A 1-inch gap between the shell and a 15-foot-long elliptically curved template when placed along the meridional surface of a shell within a single plate section and not closer than 12 inches to a welded seam.

6. A 1 1/2-inch gap when the elliptical template above was placed across one or more welded seams.

i l

l 6

l A e :iAGiG B. 5-4 l

l -- . - _ _ , . - _ _ , _ . _ _ . _ _ _ _ _ , _ . , . . _ . , . _ _ . , . _ _ , _ _ . . . . _._. _ ,, _ _ _ _ , , _ , . , _ , _

B/B-FSAR AMENDMENT 42 MAY 1983 B.6 CONTAINMENT STEEL BOUNDARY NOT BACKED BY CONCRETE i The materials, f abrication, installation and testing requirements were in accordance with the 1971 ASME B&PV Code,Section III, Division 1, Subsection NE, with Addenda through Summer 1973.

B. 6.1 Materials The materials complied with the requirements of the 1971 ASME l BSPV Code,Section III, Division 1, Paragraph NE-2000, and also i to the following specifications:

APPLICATION SPECIFICATION Emergency Personnel Airlock and Equipment Access Hatch with Integral Personnel Airlock SA516 Grade 70 Penetration Pipe Sleeves j (i) up to 24 inch diameter SA-333 Grade 1 or 6 Seamless (ii) over 24 inch diameter SA-516 Grade 60 B.6.2 Ouality control B.6.2.1. Testing B.6.2.1.1 General The testing of the containment leaktight boundaries not backed by concrete were in accordance with the ASME B&PV Code,Section III, Division I, Subsection NE-5000.

! B.6.2.1.2 Testino of Welds One hundred percent of all welds between penetration and flued fitting, and flued fittings and pipelines were examined by radiographic examinations. One hundred percent of all welds in l

i the equipment hatch, personnel airlock, and penetration sleeves were inspected also by radiographic examination where possible.

! Where radiography could not be employed, ultrasonic examination was used. Penetration to insert plate welds and penetration to liner welds were magnetic particle or liquid penetrant examined 1 in lieu of 1005 radiography. Penetration insert plate to liner

! weld was spot radiographed and magnetic particle or liquid penetrant examined in lieu of 1005 radiography. Penetration ir. sert plate to frame welds for air locks and access openings were magnetic particle examined or liquid penetrant examined in lieu of 1005 radiography. If a weld failed to meet the acceptance standards specified in NE-5300,Section III of the ASME B6PV Code, the entire unacceptable weld was removed and the joint rewelded. Ihe repaired areas were radiographed.

B.6-1 AU146.16

. . i

" B/B-FSAR AMENDMENT 42 l MAY 1983 B.6.2.2 Fabrication and Installation B.6.2.2.1 General The fabrication and installation of the containment steel boundaries not backed by concrete were in accordance with the ASME B&PV Code,Section III, Division I, Subsection NE-4000.

B.6.2.2.2 Oualification of Welders The qualifications of welders and welding procedures were in accordance with Section III, Division 1, Subsection NE-4300 of the ASME B&PV Code. l i

l l

i B. 6- 2 l

1

B/B-FSAR AMENDMENT 43 SEPTEMBER 1983 B.7 STAINLESS STEEL POOL LINERS The liner for the spent fuel pool, fuel transfer canal and spent f uel cask pit are not covered by this section. For f urther details on these liners refer to Subsection 9.1.2. 3.

B.7.1 Materials Stainless steel pool liners were fabricated from A240 Type 304 material, hot rolled, annealed and pickled and furthec processed by cold rolling.

B.7.2 Welding Welding procedures were in accordance with the ASME B&PV Code,Section III, Division 2, Paragraph CC-4540, and ASME Section IX. All seam welds were complete penetration groove square butt welds.

The liner plate seam welds were examined and tested as follows:

a. Radioagraphic examination was per formed in accor-dance with the requirements of ASME Section V,

" Nondestructive Examination." A minimum of 2%

of all liner seam welds were examined.

b. Ultrasonic examination may be performed in lieu of radiography on liner seam welds when joint detail does not permit radiographic examination.

c. Liquid penetrant examination was performed on austenitic materials. The weld surfaces and at least 1/2 inch of the adjacent base material on each side of the weld were examined. The examination coverage was 100% of all shop and field seam welds,

d. Vacuuta leak test was per formed for leaktightness on all liner plate seam welds.

B.7.3 Erection Tolerances Tolerances for free-standing liner work conformed to CC-4522.1.1 of Section III, Division 2 of ASME with the following additional requirements for the refueling water storage tanks:

B.7-1 A0014618 l

L

= . _ . . . . . - ._ - . .- .

I AMENDMENT 43 l B/B-FSAR SEPTEMBER 1983

a. The radius of the cylindrical shell was within

+ 3 inches of the theoretical radius. Radius

measurements were made and at 36' increments at 10 foot incrementsverticallyl circumferential1y.

9

b. The railus of the inner surf ace of the dome does not deviate from the design value by more than l

+ 3 inches. The height of the dome above the '

i spring line was not greater than 6 inches above j

the design height, and in no case was it lesc than the design height above the spring line.

1 1

I I

3 l

j l

i 1

A i

j l

}

B.7-2 AUU3AC19 l

E/E-FSAP B. 9 OTHER STAINLESS STEEL ELEMENTS Stainless steel embedded plates and stainless steel checkered floor plates were fabricated from A240 Type 304 material, hot rolled, annealed and pickled.

Stainless steel bars and rounds were fabricated from A276 or A479 Type 304 material, hot rolled, annealed and pickled.

Stainless steel pipes were fabricated from A312 Type 304 or A358 Type 304 or A376 Type 304 materials, hot rolled, annealed and pickled.

Stainless steel gratings were fabricated from A240 Type 302 or Type 304 materials, hot rolled, annealed and pickled prior to fabrication and then electropolishei after fabrication.

Stainless steel sump liners were fabricated from A240 Type 304 or Type 316 materials.

stainless steel bolts were fabricated from A 193 Type 304 Class 1 material.

Stainless steel nuts were f abricated from A194 Type 304 material.

Stainless shapes were fabricated from A276 or A479 Type 304 materials.

For further discussion on austenitic stainless steel, refer to subs ection 5. 2. 3. 4.

AWMGZO E.e-1

B/B-FSAR AMENEMENT 19 MARCH 1979 B.9 NUCLEAR STEAM SUPPLY SYSTEM (NSSS) COMPONENT SUPPORT STEEL B. 9.1 General

! Material and Quality control Programs for component support steel ,

l conformed to the requirements of Subsection NF of the 1974 ASME Code, Summer of 1975 addenda,Section III, Division I. All further references to Subsection NF in this section on NSSS ,

! component supports imply the same edition and addenda. l

B.A.2 steel Materials Component support steel materials are summarized in Table B.9-1.

B.9.3 Weldino cualifications -

All welding procedures were qualified in accordance with the welding procedure qualification requirements of NF-4300 of ASME Section III, Subsection NF.

B.9.4 ouality Control i

B.9.4.1 General Certified material test reports which provide the results of all chemical analyses and mechanical tests were furnished in accordance with the requirements of NF-2000. Test reports d

included the results of Charpy Impact Tests which conformed to j Subsection NF of the ASME Code. Identification of material .

i requiring traceability was provided in compliance with Section

, III of the ASME Code.

B.9.4.2 Lamination Tests Plates loaded in tension during service in the through thickness (short-transverse) direction, a s defined in NF-3226.5, Subsection NF of ASME,Section III, were examined by the straight beam ultrasonic method in accordance with ASME SA-578.

B.9.4.3 Nond estructive Examination of Welds Nondestructivt examinations were conducted in accordance with the requirements of ASME Section V and NF-5000 of Section III Subsection NF. Acceptance standards for radiography, ultrasonic, magnetic particle, liquid penetrant, and visual examinations, complied with the requirements of NF-5000 of Section III, Subsection NF.

j B.9.5 Fabrication and Installation i

The fabrication and installation of NSSS component supports were accomplished in conformity with NF-4000 of ASME Section III, Subsection NF.

A003A621 B. 9-1 l

B/B-FSAR B. 9. 5.1 Installation Tolerances Installation tolerances for (a) NSSS component support embedment location, and (b) centerlines and work points with reference to inplace NSSS component supports are as follows;

a. NSSS component support embedment location tolerances:

work point elevation deviation... i 1/2 inch horizontal deviation.. i 1/2 inch embedment axis angula r devia tion.. . . . i 30 minutes

b. Tolerances for centerlines and workpoints with reference to inplace NSSS component supports:

work point elevation deviation. . . i 1/2 inch horizontal deviation.. i 1/2 inch centerline and angular deviation..... i 15 axis minutes deviation from true line.................. i 1/2 inch I

B.9-2 AOO3.4622 L

m ,

=0 I $

B/B-FSAR AMENDMENT 45 JUNE 1984 TABLE B.9-1 MATERIAL FOR NSSS COMPONENT SUPPORTS MATERIAL APPLICABLE ASME SPECIFICATION NUMBER PRODUCT FORM CODE PROVISION A618 GRADE III TUBE CODE CASE 1644 A588 GRADE A, 8 PLATES, BARS CODE CASE 1644 SHAPES SA-540 GR. B24 CLASS 1 BOLTING MATERIAL SECTION III AND CLASS 4 SUBSECTION NA TABLE I-13.3 A490 BOLTING MATERIAL CODE CASE 1644 SA-194 GR 7 NUTS SUBSECTION NA TABLE I-13.3 SA-533 CLASS 2 PLATE SUBSECTION NA TABLE I-1.1 l

l l

l l B.9-3 l

US A0014623

• CE-1-A Tspical Raoort s

l 4 ,

QUALITY ASSURANCE PROGRAM FOR NUCLEAR GENERATING STATIONS l

Commonwealti. Edison Cosipany vs.:

.D6a A0014624

4

2. QUALITY ASSURANCE PROGRAM k.1 General The comonwealth Edison Company has extensive experience with the development, scheduling, design, construction and operation of electric generating facilities. It has pioneered in comercial nuclear power. Comonwealth Edison Company and its consultants and vcndors have~ established designs and specifications for compliance with applicable regulations, ASME Code and National Standards to casure installations of utmost safety and reliability.

Comonwealth Edison Company has attained qualified equip-m:nt vendors and contractors through experience, evaluation at vcador plants and site surveillance during plant erection. These cfforts provide assurance that compliance with applicable design cpecifications and codes is maintained and a high level of roliability is achieved.

2.2 Policy It' is the policy of Comonwealth Edison Company to assure a tigh degree of functional integrity of the equipment, structures and

! safety-related systems of its nuclear generating facilities, the p3rformance of which are essential to the prevention of nuclear cccidents that could cause undue risk to the health and safety of -

the public, or to the mitigation of the consequences of such cccidents in the unlikely event they should occur. These systems, equipment, and structures will be identified, designed, fabricated, crected, tested and operated to the criteria and requirements of ASME Boiler and Pressure Vessel Code Sections III and XI, as rcforenced in the SAR applicable to a specific unit at the time of Cngineering, construction and major modifications, Appendix B to 10CFR50, Subpart M to 10CFR71, and the mandatory requirements of *32 t

ANSI Il45.2 and N18.7 Standards. Also, Commonwealth Edison commits to comply with 10CFR Part 21 " Reporting of Defects and Nancompliance" and with the Regulatory position of the following Regulatory Guides and the requirements of the following ANSI Standards as listed for each station.

i 1 2-1 l

l AOOl%'26 l

Rev. 32 10-05-84

. Commonwealth Edloon

4 I. Dresden, Quad Cities and Zion Nuclear Power Stations 1.28 (Safety Guide 28 - 6/72): 1.30 (Safety Guide 30 8/72):

1.33 (Safety Guide 33 - 11/72): 1.37 - 3/73: 1.38 - 3/73: 1.39

- 3/73: 1.54 - 6/73: 1.58 - Rev. 1; 1.64 - 10/73: 1.74 - 2/74:

1.8 (Safety Guide 8 - 3/71); 1.146 - Rev. 0 8/80: ANSI N45.2.8-74 (Draft 3, Rev. 3, N45.2.9-74 (Draft 15, Rev. O, 4/74); ANSI N45.2.12-74 (Draft 3, Rev. 4 2/74); ANSI N45.2.13-74 (Draft 2, Rev. 4, 4/74)

II. LaSalle County Nuclear Power Station -

1.28 - Rev. O, 6/72; 1.30 - Rev. O, 8/72: 1.33 - Rev. 2; 1.37

- Rev. 0, 3/73: 1.38 - Rev. 2; 1.39 - Rev. 2; 1.54 - 6/73; 1.58 - Rev. 1; 1.64 - Rev. 2; 1.74 - Rev. O, 2/74: 1.8 - Rev.

1-R; 1.88 - Rev. 2; 1.94 - Rev. 1; 1.116 - Rev. 0-R, 6/76;

! 1.123 - Rev. 1; 1.146 - Rev. O, 8/80; 1.144 - Rev. 1

! III. Byron and Braidwood Nuclear Power Stations 1.28 - Rev.1; 1.30 - Rev. O, 8/72; 1.33 - Rev. 2: 1.37 Rev. O, 3/73; 1.38 - Rev. 2; 1.39 - Rev. 2; 1.54 - 6/73; 1.58 - Rev.

1: 1.64 - Rev. 2; 1.74 - Rev. O, 2/74; 1.8 Rev. 1-R; 1.88 -

Rev. 2; 1.94 - Rev. 1; 1.116 R, 6/76; 1.123 - Rev. 1; 1.146 Rev. O, 8/80; 1.144 - Rev. 1 i

Exceptions or alternatives to this Topical Report for specific l

plants identified in the Safety Analysis Report or Technical Specification will take precedence over commitments in this Topical Report.

It is also the policy of Commonwealth Edison Company to l

cssure a high degree of functional integrity for its generating facilities so as to achieve high availability of these facilities for the production of electrical power and to maintain overall quality levels which will achieve the foregoing in a safe, effective and economic manner.

The Quality Requirements and Quality Procedures of the Company Quality Assurance Program Manual (see Appendix A for Program Manual Index) as described herein, document the written policies and procedures of the Program and are augmented by other written Depart-

• ment and Station procedures and instructions. In the Quality

• Assurance Department controlled Quality Assurance Department

• instructive, directive and procedural t y documents are used to further explain in specific detail certain department activities as **

deemed needed. Such documents may cover items such as training, e training program, review of procurement documents, personnel qualification and certification, maintenance and updating of ASME #29 Code information on the Approved Bidders List, etc. In addition, Site Quality Assurance Departments may have controlled Site l

instructions to provide detailed explanations and methodologies for **

I implementing the Quality Requirements and Quality Procedures, and to

• i provide instructious for other site quality activities as necessary.

2-2 g dWM626 j CommonweeMh Edioen

I

. i.

TEo Coggg October 9, 1985 40 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD $[5 In the Matter Of: ) 35 0CI15 N028

)

COMMONWEALTH EDISON COMPANY ) -

) Docket Nos. 50-4560L-$$Q'j,'y , ,

50-457 (Braidwood Station, Units 1 ) BRANCH and 2) )

CERTIFICATE OF SERVICE I, Rebecca J. Lauer, one of the attorneys for Commonwealth Edison Company, certify that the following persons have been served in the above-entitled matter with copies of the documents indicated and that service has been executed in the manner indicated.

• • • • Lawrence Brenne r, Esq. ***br. A. Dixon Callihan Chairman Administrative Law Judge Administrative Law Judge 102 Gak Lane Atomic Safety nnd Licensing Oak Ridge, TN 37830 Board United States Nuclear Regulatory **Stuart Treby, Esq.

Commission Elaine I. Chan, Esq.

Washington, DC 20555 Office of the Executive Legal Director

• • • Herbert Grossman, Esq. United States Nuclear Regulatory Administrative Law Judge Commission Atomic Safety and Licensing Washington, DC 20555 Board United States Nuclear Regulatory ***htomic Safety and Licensing Commission Board Panel Washington, DC 20555 United States Nuclear Regulatory Commission

• • • • Dr. Richard F. Cole Washington, DC 20555 Administrative Law Judge Atomic Safety and Licensing Board United States Nuclear Regulatory Commission Washington, DC 20555 2

. \

• • • %tomic Safety and Licensing * * **Charle s Jones , Director Appeal Board Panel Illinois Emergency Services United States Nuclear Regulatory and Disaster Agency Commission 110 East Adams Washington, DC 20555 Springfield, IL 62705

• • • Mr. William L. Clements **Hilliam Little, Director Chief, Docketing and Services Braidwood Project United States Nuclear Regulatory Region III Commission United States Nuclear Regulatory Office of the Secretary Commission Warnington, DC 20555 799 Roosevelt Road Glen Ellyn, IL 60137

• • • • Ms. Bridget Little Rorem 117 North Linden Street ** Dan Stevens P.O. Box 208 United States Nuclear Regulatory Essex, IL 60935 Commission 7920 Norfolk Avenue

• Robert Guild Phillips Building Douglass W. Cassel, Jr. Bethesda, MD 20014 Timothy W. Wright, III BPI 109 North Dearborn Street Suite 1300 Chicago, IL 60602

• Messenger delivery on October 9, 1985 of the following documents:

1) Applicant's Tenth Partial Response to Rorcm's First Set of Quality Assurance Interrogatories and Request to Produce dated October 9, 1985.

2) Affidavits of Michael J. Wallace, Charles M.

Allen, Eugene E. Fitzpatrick, Louis O. Del George, and Cordell Reed.

3) List of documents withheld by Applicant under a claim of privilege.

4) Documents referenced in the response to specific interrogatory 6.

5) Lists of home addresses and telephone numbers of individuals identified in responses to specific interrogatories 19, 51, 58, and 59.

i

_-___-______J

r 9

• • Deposit in the United States mail on October 9, 1985 the following documents:

1) Applicant's Tenth Partial Response to Rorem's First Set of Quality Assurance Interrogatories and Request to Produce dated October 9, 1985.

2) Affidavits of Michael J. Wallace, Charles M.

Allen, Eugene E. Fitzpatrick, Louis O. Del George, and Cordell Reed.

3) List of documents withheld by Applicant under a claim of privilege.

4) Documents referenced in the response to specific interrogatory 6.

5) Lists of home addresses and telephone numbers of individuals identified in responses to specific interrogatories 19, 51, 58, and 59.

• • • Deposit in the United States mail on October 9, 1985 the following documents:

1) Applicant's Tenth Partial Response to Rorem's First Set of Quality Assurance Interrogatories and Request to Produce dated October 9, 1985.

2) Affidavits of Michael J. Wallace, Charles M.

Allen, Eugene E. Fitzpatrick, Louis O. Del George, and Cordell Reed.

3) List of documents withheld by Applicant under a claim of privilege.

4) Documents referenced in the response to specific interrogatory 6.

• • • • Deposit in the United States mail on October 9, 1985 the following documents:

1) Applicant's Tenth Partial Response to Rorem's First Set of Quality Assurance Interrogatories and Request to Produce dated October 9, 1985.

2) Affidavits of Michael J. Wallace, Charles M.

Allen, Eugene E. Fitzpatrick, Louis O. Del George, and Cordell Reed.

l L d

c- ,

, 'i

3) List of documents withheld by Applicant under a claim of privilege.

Rebeccf J. Lauer ISHAM, LINCOLN & BEALE Three First National Plaza Suite 5200 Chicago, Illinois 60602 (312) 558-7500 DATED: October 9, 1985

-4 I